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Sample records for bwr stability operating

  1. Valuation of BWR stability operating in natural circulation conditions

    International Nuclear Information System (INIS)

    Nowadays, the design of reactors having appropriate stability margins, the adoption of operating procedures avoiding possible unstable regions and the development of mitigation strategies to cope with inadvertent instability occurrences have strongly limited safety concerns in this regard. However, despite the obvious need for plant-specific Probabilistic Safety Assessment (PSA), BWR (boiling water reactor) transients of general interest can be identified and characterized as for example, overpressurization events, Large Break Loss of Coolant Accidents (LBLOCAs), feedwater temperature decrease, increase of core flow, main circulation pump flow rate increase, control rod withdrawal and others. Simulations of these complex scenarios have been improved by the utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes. In this work, the RELAP5/MOD3.3 thermal-hydraulic system code and the PARCS/2.4 3D neutron kinetic code have been adopted to predict the Peach Bottom BWR stability during recirculation pumps trip while the reactor is operating in a special region of power and core flow map. In the recirculation pump trip event, the stopping of the recirculation pumps causes a sharp decrease in the core flow, which generates a considerable negative reactivity insertion that tends to reduce power and, consequently, the amount of steam generated. The BWR reactor stability has been valuated during natural circulation conditions after the pumps trip event. The time evolution of the power and the related thermal-hydraulic parameters were investigated to analyse the behavior of the reactor for this special operation condition. (author)

  2. BWR stability analysis

    International Nuclear Information System (INIS)

    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

  3. Improvement for BWR operator training

    International Nuclear Information System (INIS)

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

  4. Long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Considering next generation requirement for nuclear plants, a long cycle operating simplified BWR (LSBWR) concept is proposed. The major features of LSBWR are; 1) Long cycle operation core using uranium fuels; 2) Simplified system and component as well as passive systems; 3) Combined building concept with ship hull structure. This concept have potential to reduce construction cost and to Increase availability. Safety feature of LSBWR makes possible to attain no evacuation capability in case of a severe accident. Further research and development is underway. (author)

  5. BWR stability using a reducing dynamical model

    International Nuclear Information System (INIS)

    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)

  6. BWR stability using a reduced dynamical model

    International Nuclear Information System (INIS)

    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

  7. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    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.)

  8. Advanced methods for BWR transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  9. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

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

  10. Impact of Plant Noise on BWR Stability Analyses

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  12. BWR stability analyses at BNL

    International Nuclear Information System (INIS)

    The March 9, 1988 instability at the LaSalle County-2 boiling water reactor power plant at Seneca, IL was simulated with Brookhaven National Laboratory's (BNL's) Engineering Plant Analyzer (EPA) for the purpose of demonstrating that the EPA is suitable for simulating large-amplitude, limit-cycle power and flow oscillations. It was shown in fall of 1988, by comparing all the available plant data from the STARTREC recording system of LaSalle-2 with EPA simulation results, that the EPA reproduces the LaSalle-2 oscillations without the use of stabilizing or destabilizing model or parameter modifications. The power vs. flow map of the LaSalle-2 plant was also reproduced at five lines of constant control rod positions. The LaSalle-2 stability boundary was established with the EPA and confirmed within ±15% accuracy by comparing the EPA results with the results of the frequency domain code LAPUR of Oak Ridge National Laboratory. Comparisons of EPA simulation results with plant data from three Peach Bottom stability tests show an agreement, based on mean and standard deviation, of -10±28%, -1±40% and +28±52% (low power) in the gain of the pressure to power transfer functions. This demonstrates that the time domain code HIPA in the EPA is capable of simulating instabilities

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

    International Nuclear Information System (INIS)

    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

  14. BWR stability analysis: methodology of the stability analysis and results of PSI for the NEA/NCR benchmark task

    International Nuclear Information System (INIS)

    The report describes the PSI stability analysis methodology and the validation of this methodology based on the international OECD/NEA BWR stability benchmark task. In the frame of this work, the stability properties of some operation points of the NPP Ringhals 1 have been analysed and compared with the experimental results. (author) figs., tabs., 45 refs

  15. 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)

  16. BWR stability analysis with the BNL Engineering Plant Analyzer

    International Nuclear Information System (INIS)

    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

  17. Space–time convergence analysis on BWR stability using TRACE/PARCS

    International Nuclear Information System (INIS)

    Highlights: ► Quantify TRACE/PARCS space–time discretization error for BWR stability prediction. ► Establish space and time discretization necessary for space–time converged models. ► Show that the space–time converged model gives more reliable results for both stable and unstable reactor. ► Use of the space–time converged model increases confidence in the prediction of BWR stability. -- Abstract: Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Even though BWR instability is not a severe safety concern, it could cause reactor scram and significantly decrease the economic performance of the plant. This paper aims to (a) quantify TRACE/PARCS space–time discretization error for simulation of BWR stability, (b) establish space (nodalization) and time discretization necessary for space–time converged model and (c) show that the space–time converged model gives more reliable results for both stable and unstable reactor. The space–time converged model is obtained when further refinement of numerical discretization parameters (nodalization and time step) has negligible effect on the solution. The study is significant because performing a space–time convergence analysis is a necessary step of qualification of the TRACE/PARCS model, and use of the space–time converged model increases confidence in the prediction of BWR stability.

  18. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    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.)

  19. LAPUR-K BWR stability benchmark

    International Nuclear Information System (INIS)

    This paper documents the stability benchmark of the LAPUR-K code using the measurements taken at the Ringhals Unit 1 plant over four cycles of operation. This benchmark was undertaken to demonstrate the ability of LAPUR-K to calculate the decay ratios for both core-wide and regional mode oscillations. This benchmark contributes significantly to assuring that LAPUR-K can be used to define the exclusion region for the Monticello Plant in response to recent US Nuclear Regulatory Commission notices concerning oscillation observed at Boiling Water Reactor plants. Stability is part of Northern States Power Reload Safety Evaluation of the Monticello Plant

  20. The stability analysis of LAPUR6 for Chinshan BWR nuclear power plant

    International Nuclear Information System (INIS)

    Chinshan Nuclear Power Plant (NPP) is the first NPP in Taiwan which is a BWR/4 plant. The original rated power for each unit was 1775 MWt. After the project of measurement uncertainty recovery (MUR) for Chinshan NPP, the operating power is 1805 MWt now. The Chinshan NPP Unit 2 cycle 23 stability analyses were performed by the LAPUR6 stability analysis methodology. Comparing the LAPUR6 stability analysis results and vendor's results, they are similar. (author)

  1. BWR core stability analysis with RETRAN

    International Nuclear Information System (INIS)

    A RETRAN model was developed for determining the stability of boiling water reactor. This model was benchmarked against plant data from stability tests conducted during plant operations. The stability analysis with RETRAN is demonstrated using best estimate RETRAN input data representative of the NSSS. All of the important neutronic and thermal hydraulic feedback mechanisms are taken into account through the modeling of the reactor vessel, recirculation loops and core neutronics. The analysis was performed with the RETRAN02 MOD003 computer code. The transient is initialized by a small step decrease in the steam dome pressure. The core exit (upper plenum) pressure and core power transient responses to this perturbation are transformed into frequency data and a system transfer function is then obtained. The system transfer function is fitted to a second order equation from which the decay ratio and natural frequency can be determined

  2. BWR operator training for emergency conditions using simulator in Japan

    International Nuclear Information System (INIS)

    BWR Operator Training Center Corporation (BTC) is the only training organization for all Japanese BWR utilities. BWR operator training program is constructed to maintain and upgrade the ability of the individual operator and the operation crews. The training course for the individual operator is focused on the operator's positions, from that of the main operator to the licensed shift supervisor. The target of the training for operation crews, which is called the family training, is to upgrade the team performance including shift supervisor. The training programs are combined to improve and upgrade the operator performance using simulator. BWR training simulator has been designed to demonstrate many kinds of malfunctions, modeling a single failure and multiple failure, including loss of coolant accident. It has also the alteration and scheduling function to simulate the abnormal and accident situation. These functions are utilized to easily simulate the accident scenario and enlarge the operator's knowledge in case of the abnormal condition. This paper describes the BWR operator training for emergency conditions using simulator and the advanced control board simulator performance including the abnormal situation functions. (author). 5 refs, 2 tabs

  3. Thermohydraulic stability coupled to the neutronic in a BWR

    International Nuclear Information System (INIS)

    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

  4. Predictions by the proper orthogonal decomposition reduced order methodology regarding non-linear BWR stability

    International Nuclear Information System (INIS)

    Unexpected non-linear boiling water reactor (BWR) instability events in various plants, e.g. LaSalle II in 1988 and Oskarshamn II in 1990 amongst others, emphasize the major safety relevance and the existence of parameter regions with unstable behavior. A detailed description of the complete dynamical non-linear behavior is of paramount importance for BWR operation. An extension of state-of-the-art methodology towards a more general stability description, also applicable in the non-linear region, could lead to a deeper understanding of non-linear BWR stability phenomena. With the intention of a full non-linear stability analysis of the two-phase BWR system, the present paper aims at a general non-linear methodology capable to achieve reliable and numerical stable reduced order models (ROMs), representing the dynamical behavior of an original system based on a small number of transients. Model-specific options and aspects of the proposed methodology are focused on and illustrated by means of a strongly non-linear dynamical system showing complex oscillating behavior. Prediction capability of the proposed methodology is also addressed. (orig.)

  5. BWR stability monitoring using adaptive methods

    International Nuclear Information System (INIS)

    Adaptive methods are system identification methods which are commonly used to analyze nonlinear and nonstationary signals. The use of three of these methods to study the time dependence evolution of the linear stability parameters of a Boiling Water Reactor using neutronic noise signals has been analyzed. Both simulated and real neutronic signals have been considered to perform the analysis

  6. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    The BWR control rods made by ABB use boron carbide (B4C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B4C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  7. Present status and subjects on the evaluation of neutronic and thermalhydraulic stability in BWR

    International Nuclear Information System (INIS)

    The boiling water reactors (BWRs) have feasibility to form flow instability and nuclear thermal coupling instability accompanied with formation of steam voids at their cores. These neutronic and thermalhydraulic stability (NTS) was regarded to one of important problems at beginning of BWR development, to carry out its research and development. As seen in the safe design judgement standard, as even when forming output oscillation not so as to exceed allowed design limit of fuels, it is requirement to easily enable to control it, for design on this stability a standard recognized that within usual operating regions is regarded to be stable is established, and it has been allowed to evaluate it by analysis. In this report, in Chapter 2 was shown what phenomenon was NTS and relationship between progress of BWR design and stability, in Chapter 3 was shown basic models and analysis method, at present, used for its stability analysis method, in Chapter 4 was described analysis codes and evaluation calculation actually used for stability analysis in Japan, at present, in Chapter 5 was described stability tests used in Japan, in Chapter 6 was described stability monitors reflecting recent researches, and in Chapter 7 was shown recent evaluation standard on channel hydraulic stability, reactor core stability and regional stability. (G.K.)

  8. Operation method for BWR type reactor

    International Nuclear Information System (INIS)

    In a BWR type reactor, the number of fuels at low enrichment, among initially loaded fuels, is increased greater than that of fuels to be exchanged, and the number of fuels at low enrichment remained in a reactor core after fuel exchange is decreased to smaller than that of entire control rods. Further, the fuels at low enrichment are disposed to the inner side except for the outermost circumference in the reactor core after fuel exchange. Since fuels of high reactivity are disposed at the outermost circumference in a second cycle, leakage of neutrons is increased and effective breeding factor is decreased. However, since the number of brought over fuels at low enrichment is decreased and the number of fuels at high enrichment is increased, effective average reactor core enrichment degree is increased, to compensate the lowering thereof due to the increase of neutron leakage. Since dispersion effect for the distribution of the enrichment degree can be utilized as much as possible by greatly reducing the number and the enrichment degree of fuels at low enrichment for initially loaded fuels, irrespective of the average enrichment degree and the fueling pattern in a first cycle, a burnup degree upon take-out of initially loaded fuels at ow enrichment degree can be increased to maximum. (N.H.)

  9. Ranking of input parameters importance for BWR stability based on Ringhals-1

    International Nuclear Information System (INIS)

    Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Uncertainty calculations for BWR stability, based on the Wilks' formula, have been already done for the Ringhals-1 benchmark. In this work, these calculations have been used to identify and rank the most important parameters affecting the stability of the Ringhals-1 plant. The ranking has been done in two different ways and a comparison of these two methods has been demonstrated. Results show that the methods provide different, but meaningful evaluations of the ranking. (author)

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

    International Nuclear Information System (INIS)

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

  11. BWR stability analysis with three-dimensional transient code

    International Nuclear Information System (INIS)

    Recently, neutron flux oscillations of two different modes were observed in several foreign BWR plants. One is core wide oscillation mode which is characterized by a phenomenon that neutron flux oscillates in-phase over a whole core. At La Salle 2 plant (U.S.A.), the amplitude of core wide neutron flux oscillation grew considerably large to result in a reactor scram, which aroused great concern about BWR stability. The other is regional oscillation mode which is characterized by the phenomenon, as typically observed at Caorso plant (Italy), that neutron flux of a half core oscillates out-of-phase to that of the other half core. These neutron flux oscillation phenomena were caused by nuclear-thermal hydraulic coupled instability and requires an evaluation study on oscillation detectability and effect on fuel integrity. Particularly, the regional oscillation mode requires three-dimensional analysis since it may bring about locally large amplitude power oscillation. For this reason, analysis was done with the three-dimensional transient code TOSDYN-2 to study reactor condition which causes the regional oscillation and also to evaluate fuel thermal margin under the neutron flux oscillations of these two instability modes. (author)

  12. BWR parametric sensitivity effect of regional mode instability on stability boundary

    International Nuclear Information System (INIS)

    Both in-phase (core wide mode) instability and out-of-phase (regional mode) instability are of great concerns in BWR stability issues. Normally, decay ratios for regional mode oscillations are much less than those under core wide conditions. This research presents a study of fractional change of decay ratio to evaluate parametric effects of regional mode instability on reload core design power/flow stability boundary for the Chinshan Nuclear Power Plant Unit 2 Cycle 21 (BWR4). Making use of LAPUR5 and SIMULATE-3 codes, we have established a methodology to conduct such out-of-phase stability analysis. Many important parameters, such as system pressure, core flow rates, coolant inlet subcoolings, and moderator void fraction, fuel physical and geometrical properties, have strong influences on regional mode stability. Current investigations have shown that at some operation points along the stability boundary, certain parameters are sensitive enough so that the regional mode decay ratio could become greater than the core wide mode decay ratio. (author)

  13. BWR parametric sensitivity effect of regional mode instability on stability boundary

    International Nuclear Information System (INIS)

    Both in-phase (core wide mode) instability and out-of-phase (regional mode) instability are of great concerns in BWR stability issues. Normally, decay ratios for regional mode oscillations are much less than those under core wide conditions. However, under certain observation mode, the regional mode instability has the phenomenon of power increasing in one half of the core and at the same time, it decrease in the other half, so it looks like that the average power remains essentially constant. This research presents a study of fractional change of decay ratio to evaluate parametric effects of regional mode instability on reload core design power/flow stability boundary for the Chinshan Nuclear Power Plant Unit 2 Cycle 21 (BWR4). Making use of LAPUR5.2 and SIMULATE-3 codes, we have established a methodology to conduct such out-of-phase stability analysis. Many important parameters, such as system pressure, core flow rates, moderator void fraction, fuel physical and geometrical properties, have strong influences on regional mode stability. Current investigations have shown that at some operation points along the stability boundary, certain parameters present more sensitive characteristics. (author)

  14. Analysis of non-linear BWR stability behavior applying proper orthogonal decomposition

    International Nuclear Information System (INIS)

    The main drivers of BWR stability behavior are the multiple thermal hydraulic interactions between power, flow rate, and density, reinforced by the Neutronics feedback. This coupling is schematically presented in Figure 1. Especially for high power low flow operating conditions associated with unfavorable power distribution BWR operation requires attention with respect to power oscillations. Admissible reactor operation conditions maintain a certain distance to the stability limit given by linear theory. Evaluation of non-linear states requires application of time domain codes or measurement data but this depends on the specific transients considered. Improvements of non-linear stability analysis focus on the accelerating of simulations and to provide assessment for the whole parameter space. In our transient analysis, the physical behavior of the system is approximated by a reduced order model (ROM) that respects stability relevant characteristics. More precisely, the system of coupled non-linear partial differential equations (PDEs) is mapped to coupled non-linear ordinary differential equations (ODEs) that can be solved faster and analyzed with respect to non-linear stability phenomena. Proper orthogonal decomposition (POD), i.e. a spectral method based on experimental or computational fluid dynamic (CFD) data, is capable to detect oscillating states of the physical system needed. Moreover, POD provides a well-defined truncation criterion for the minimum number of modes. A standard Galerkin method employing POD modes as Ansatz functions yields a non-linear ROM. The exceptional advantage of our methodology is its generality. It is accessible for various physical systems including the reactor dynamics of BWR. We envision a fully coupled non-linear investigation of the BWR system. The method benefits from a well defined sequence of processing steps which are automated to a large extent. This minimizes the required user interaction. Obviously the user still needs to

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-09-15

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

  16. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    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)

  17. Transient and stability analysis of a BWR core with thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779 Col. Narvarte, 03020 Mexico, DF (Mexico); Espinosa-Paredes, Gilberto [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, DF (Mexico)], E-mail: gepe@xanum.uam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec Mor. (Mexico)

    2008-08-15

    The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket-seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal-hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO{sub 2} core, even during transient conditions. The stability and transient analysis show that the thorium-uranium fuel can be operated safely in current BWRs.

  18. 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

  19. Recent training technology of BWR operators using full scope simulators

    International Nuclear Information System (INIS)

    Nuclear power plants are being operated at high standards now in Japan. There are far few opportunities for operators to perform in challenging situations. To maintain skill and refinement, the simulator training is indispensable for them. BWR Operator Training Center (BTC) provides training courses according to the grade and duty of the operators. The training force constitutes of personnel from utilities', manufacturers' and also BTC-hired personnel. One of the big features of BTC training is composite team type. In this form of training, men from different plants make a team and help each other study. On the human factor viewpoint, error experience on simulators is one of the important items. Training on recognizing subtle symptom is an example of a recent development. Team training for actual crew is effective from various viewpoints. (author)

  20. Development of BWR operator training simulator and training support systems

    International Nuclear Information System (INIS)

    This paper describes a BWR operator training simulator and training support systems that have been developed with the aim of providing support throughout operator training. The operator training simulator is needed in order to improve simulation fidelity and enlarge simulation scope. A 3-dimensional reactor core model has been developed in order to improve the understanding of operators respecting neutronics through realistic training. A severe accident model has been developed for training operators and technical support center teams respecting plant operation and for studying various phenomena. The severe accident is simulated by connecting the physical parameters continuously from the conventional model to the severe accident model. An emergency procedure guideline support system is adopted in order to improve efficiency of operation training for emergencies, since the emergency operation procedures are complicated and based on multiple parameter conditions. The operator training support system is also introduced so as to help training instructors to evaluate the operation and to give instructions to operators to improve operational accuracy. An instructor's burden is eased by automatically evaluating the operation errors based on signals of a simulator. The effects of these systems are evaluated and found to be effective in an actual training center and in engineers' examinations. (author)

  1. Core concept for long operating cycle simplified BWR (LSBWR)

    International Nuclear Information System (INIS)

    An innovative core concept for a long operating cycle simplified BWR (LSBWR) is currently being developed under a Toshiba Corporation and Tokyo Institute of Technology joint study. In this core concept, the combination of enriched uranium oxide fuels and loose-pitched lattice is adopted for an easy application of natural circulation. A combination of enriched gadolinium and 0.7-times sized small bundle with peripheral-positioned gadolinium rod is also adopted as a key design concept for 15-year cycle operation. Based on three-dimensional nuclear and thermal hydraulic calculation, a nuclear design for fuel bundle has been determined. Core performance has been evaluated based on this bundle design and shows that thermal performance and reactivity characteristics meet core design criteria. Additionally, a control rod operation plan for an extension of control rod life has been successfully determined. (author)

  2. Trends in BWR transient analysis

    International Nuclear Information System (INIS)

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

  3. Upgrading BWR training simulators for annual outage operation training

    International Nuclear Information System (INIS)

    Based upon the recently developed quality assurance program by the Japanese electric companies, BWR Operator Training Center (BTC) identified the needs to enhance operators' knowledge and skills for operations tasks during annual outage, and started to develop a dedicated operator training course specialized for them. In this paper, we present the total framework of the training course for annual outage operations and the associated typical three functions of our full-scope simulators specially developed and upgraded to conduct the training; namely, (1) Simulation model upgrade for the flow and temperature behavior concerning residual heat removal (RHR) system with shutdown cooling mode, (2) Addition of malfunctions for DC power supply equipment, (3) Simulation model upgrade for water filling operation for reactor pressurization (future development). We have implemented a trial of the training course by using the upgraded 800MW full-scope training simulator with functions (1) and (2) above. As the result of this trial, we are confident that the developed training course is effective for enhancing operators' knowledge and skills for operations tasks during annual outage. (author)

  4. A Simplified Spatial Model for BWR Stability Linear Accelerator and CMS Monaco Treatment Planning System

    International Nuclear Information System (INIS)

    The stability of boiling water reactors (BWRs) has been a major field of research in nuclear science ever since BWRs were in use. Since the late 1970’s, experiments have indicated the possibility for BWRs to lose their stability under certain operating conditions, characterized by a low coolant flow and high power(1–3). The source of these instabilities is the strong coupling between the neutronics and the thermalhydraulic properties of the reactor due to the significant void fraction in BWRs(4-5). To date, two kinds of such instabilities have been observed in BWR plants(4): 1. Global (corewide) oscillations, where the power produced by the fuel rods in the core oscillates uniformly in phase. 2. Regional (first azimuthal mode) oscillations, where the power in half of the core oscillates out-of-phase with respect to the other half while the average power remains essentially constant.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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, 1.0-2.0x106 kg/m2/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)

  7. Analysis of relationship between stability and flow parameters in a BWR

    International Nuclear Information System (INIS)

    Results of quantitative analysis of mutual relationship between the BWR stability and channel steam velocity are presented. The stability parameter, defined by the damping ratio, and the steam velocity are estimated by analysis of neutron noise data from local power range monitor (LPRM) detector signals. These parameters are treated as varying randomly as a function of time

  8. Experimental investigation of natural circulation BWR core-wide and regional stability on the basis of time series analysis

    International Nuclear Information System (INIS)

    A time series analysis method was performed to calculate decay ratios from dominant poles of a transfer function by applying AR method to time series of the core inlet flow rate. By utilizing this method, one can estimate stability at any specific operating point online without assuming excessively conservative conditions. Experiments were conducted with the SIRIUS facility, which simulates a representative natural circulation BWR. Channel and regional stability decay ratios at the nominal operating condition were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to investigate the effects of the design parameters on stability. For the marginal operating condition, the system further stabilized with decreasing a ratio of outer to inner power, a core inlet subcooling, and void reactivity coefficient. The system became the least stable condition when thermal conductance of the fuel rod coincided with the oscillation period of thermal-hydraulic instability. (author)

  9. Semi-automated proper orthogonal decomposition reduced order model non-linear analysis for future BWR stability

    International Nuclear Information System (INIS)

    Highlights: • Techniques within the field of ROMing based on POD are reviewed regarding “well-behaved” applications. • A systematic, general, mostly automated, reduction methodology based on POD is derived. • It is applicable for many classes of dynamical problems including the envisioned BWR application. • Robustness of this approach is demonstrated by a “pathological” test example. • The derived ROM accurately predicts dynamics of transients not included in the data set. - Abstract: Thermal–hydraulic coupling between power, flow rate and density, intensified by neutronics feedback are the main drivers determining the stability behavior of a boiling water reactor (BWR). High-power low-flow conditions in connection with unfavorable power distributions can lead the BWR system into unstable regions where power oscillations can be triggered. This important threat to operational safety requires careful analysis for proper understanding. Current design rules assure admissible operation conditions by exclusion regions determined by numerical calculations and analytical methods based on non-linear states for specific transients. Analyzing an exhaustive parameter space of the non-linear BWR system becomes feasible with methodologies based on reduced order models (ROMs), saving computational cost and improving the physical understanding. A new self-contained methodology is developed, based on the general general proper orthogonal decomposition (POD) reduction technique. It is mostly automated, applicable for generic partial differential equation (PDE) systems, and reduces them in a grid-free manner to a small ordinary differential equation (ODE) system able to capture even non-linear dynamics. This allows a much more extensive analysis of the represented physical system. Symbolic mathematical manipulations are performed automatically by Mathematica routines. A novel and general calibration roadmap is proposed which simplifies choices on specific POD

  10. Accumulation of operator workload data by using A-BWR training simulator

    International Nuclear Information System (INIS)

    Human-machine interface (HMI) of A-BWR has been developed in order to improve operational safety, reliability and to reduce workload. A-BWR HMI is fully computerized. JNES (Japan Nuclear Energy Safety Organization) and BTC (BWR Operator Training Center Corporation) have accumulated the operator workload quantitative data, related to the observation and operation at typical transient conditions, in order to evaluate the difference of operational workloads between A-BWR HMI and conventional type HMI. The workload evaluation shows the following results: - The workload density (observation and operation frequencies per unit time) of ABWR just after the plant trip is less than that of BWR-5. - At stable conditions after the transient, the workload density of ABWR becomes higher comparing that of BWR-5. A-BWR alarm system may increase the workload density caused by alarm multi-layer structure, because an operator has to use the flat and/or the CRT display to pursue every alarm. The analysis of shift team training at BTC shows that total workload is reduced at ABWR but alarm confirmation work still remains as burden. These results show some modifications might be needed for future HMI. To grasp the tendency of operator workload difference by the control panel type difference, the operator workload quantitative data have accumulated using ABWR type simulator and conventional type simulator at the same typical transient condition. These data were arranged operation frequency data, number of alarm generating data and number of switching CRT pictures data according to plant behaviour. The ABWR type HMI's characteristic have become clear by these operator workload data and the team characteristic evaluation data which BTC evaluated comparing the team performance difference of HMI type

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

    International Nuclear Information System (INIS)

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

  12. Prediction of the stability of BWR reactors during the start-up process

    International Nuclear Information System (INIS)

    The Boiling Water Reactors (BWR) are susceptible of uncertainties of power when they are operated to low flows of coolant (W) and high powers (P), being presented this situation mainly in the start-up process. The start-up process could be made but sure if the operator knew the value of the stability index Decay reason (Dr) before going up power and therefore to guarantee the stability. The power and the flow are constantly measures, the index Dr could also be considered its value in real time. The index Dr depends on the power, flow and many other values, such as, the distribution of the flow axial and radial neutronic, the temperature of the feeding water, the fraction of holes and other thermohydraulic and nuclear parameters. A simple relationship of Dr is derived leaving of the pattern reduced of March-Leuba, where three independent variables are had that are the power, the flow and a parameter that it contains the rest of the phenomenology, that is to say all the other quantities that affect the value of Dr. This relationship developed work presently and verified its prediction with data of start-up of commercial reactors could be used for the design of a practical procedure practice of start-up, what would support to the operator to prevent this type of events of uncertainty. (Author)

  13. BWR core stability prediction on-line with the computer code matstab

    International Nuclear Information System (INIS)

    MATSTAB is a computer program for three-dimensional prediction of BWR core stability in the frequency domain. This tool has been developed, and is currently used, to perform core design and optimisation with regard to core stability. The requirement regarding the predicted decay ratio of the new core is one of the limiting factors, or key parameters, in core design. To be useful, the tool should be fast and simple to apply. The results must be delivered promptly and experts should not be required to interpret them. Alternatively, the area of application for MATSTAB can be described as on-line monitoring using predictive tools. Core stability properties can be calculated for a number of presumptive reactor states, planned or unplanned. A 3-D code operating in the frequency domain may be the best tool to use for the purposes just mentioned. Some strong advantages are that the results are given promptly, they require no post-processing and are directly amenable to graphic presentation of eigenvectors, etc. (authors)

  14. Stability and vibration analyses of high power BWR using large assembly with small pins

    International Nuclear Information System (INIS)

    The Large Assembly with Small Pins (LASP) is an evolutionary fuel assembly concept to improve BWRs' economic attractiveness by increasing the core power density. It replaces four traditional assemblies with a single 22x22 large assembly. Previously published steady-state and transient analyses show that, when operating under the same power to flow ratio, the LASP core allows for 20% higher power density than the traditional fuel assemblies while maintaining the same dryout margin. In this study, mechanical vibrations and thermal-hydraulics stability analyses of the LASP core are presented. The modified Paidoussis correlation was used to calculate the vibration amplitude ratio to the fuel pin diameter and to the pin-to-pin gap of the LASP fuel and the reference fuel. The vibration ratios of the LASP are found to be higher than those of the reference core, but well below the acceptable limits. It is possible to add two spacers to the LASP assemblies to preserve the vibration ratios of the reference core. Given a more negative void reactivity coefficient than traditional BWR cores, unstable power and flow oscillations are a potential concern. Characteristics of density-wave oscillations in the LASP core and their sensitivity to operating parameters have been investigated with a linear perturbation analysis. Although the decay ratios for the LASP core are higher than those of a reference core, they are well within the traditional limits. Thus, stability is maintained in the LASP core at 20% higher reactor thermal power than the reference core. (author)

  15. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study

    International Nuclear Information System (INIS)

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  16. Parameter estimation during a transient - application to BWR stability

    Energy Technology Data Exchange (ETDEWEB)

    Tambouratzis, T. [Institute of Nuclear Technology - Radiation Protection, NCSR ' Demokritos' , Aghia Paraskevi, Athens 153 10 (Greece)]. E-mail: tatiana@ipta.demokritos.gr; Antonopoulos-Domis, M. [Institute of Nuclear Technology - Radiation Protection, NCSR ' Demokritos' , Aghia Paraskevi, Athens 153 10 (Greece)

    2004-12-01

    The estimation of system parameters is of obvious practical interest. During transient operation, these parameters are expected to change, whereby the system is rendered time-varying and classical signal processing techniques are not applicable. A novel methodology is proposed here, which combines wavelet multi-resolution analysis and selective wavelet coefficient removal with classical signal processing techniques in order to provide short-term estimates of the system parameters of interest. The use of highly overlapping time-windows further monitors the gradual changes in system parameter values. The potential of the proposed methodology is demonstrated with numerical experiments for the problem of stability evaluation of boiling water reactors during a transient.

  17. Parameter estimation during a transient - application to BWR stability

    International Nuclear Information System (INIS)

    The estimation of system parameters is of obvious practical interest. During transient operation, these parameters are expected to change, whereby the system is rendered time-varying and classical signal processing techniques are not applicable. A novel methodology is proposed here, which combines wavelet multi-resolution analysis and selective wavelet coefficient removal with classical signal processing techniques in order to provide short-term estimates of the system parameters of interest. The use of highly overlapping time-windows further monitors the gradual changes in system parameter values. The potential of the proposed methodology is demonstrated with numerical experiments for the problem of stability evaluation of boiling water reactors during a transient

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

    International Nuclear Information System (INIS)

    for a large reactor could reach 20 gigabytes) that it is not possible to load into RAM memory of an operating system with 32 bit architecture. A special procedure has been developed within the MATLAB environment to remove this memory limitation, and to invert such large matrices and finally obtain the reactor transfer functions that enable the study of system stability. Various applications of the present frequency-domain code to a typical BWR fuel assembly, a BWR core, and to a chemical reactor showed a good agreement with reference results. (author)

  19. BWR core thermal-hydraulic stability in anticipated transients without SCRAM

    International Nuclear Information System (INIS)

    The potential exists for core thermal hydraulic density wave oscillations to occur in the boiling water reactor (BWR) during anticipated transients without scram (ATWS) events. Conditions which may lead to oscillations in ATWS result either from recirculation pump trip (RPT) from appropriate ATWS signals, or from the failure to scram during core oscillations. A program to evaluate these scenarios will assess the adequacy of the system design and emergency operating procedures relative to the reactor system performance. Evaluation of ATWS events with oscillations is extremely complex, and defining the bases for the analyses is critical to the success of the program. Accurate analysis requires detailed modeling of the sequence of events, the plant configuration, and initial conditions. In conclusion: The likelihood and magnitude of any core thermal-hydraulic oscillations for a given ATWS event sequence depend not only on the choice of initial reactor conditions and fuel characteristics, but also on system dynamics and operator actions. Because of the complexity of the phenomena, accurate analysis requires detailed modeling of all relevant systems and interactions. In addition, selection of parameters must be consistent both with the intended application of the results and across the full spectrum of input conditions. Analysis of ATWS/stability events requires a reasonably limiting but representative set of plant conditions and accurate simulation of operator actions. (authors)

  20. Analysis of the OECD/NRC Oskarshamn-2 BWR stability benchmark

    International Nuclear Information System (INIS)

    Highlights: • The reference solution for Oskarshamn-2 BWR stability benchmark. • Description of TRACE/PARCS modeling of Oskarshamn-2. • Description of FW temperature correction and its effect on the calculated results. • Description of space-time convergence discretization and its effect on the calculated results. - Abstract: On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations. The event was successfully modeled by the TRACE/PARCS coupled system code, and further uncertainty analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations, and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal–hydraulics (TH), and TH/NK coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one

  1. Prediction of the stability of BWR reactors during the start-up process; Prediccion de la estabilidad de reactores BWR durante el proceso de arranque

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz E, J.A.; Castillo D, R. [ININ, Km. 36.5 Carretera Mexico-Toluca, 52045 Salazar, Estado de Mexico (Mexico); Blazquez M, J.B. [Centro de Investigaciones Energetics, Medioambientales y Tecnologicas, Av Complutense 22, 28040 Madrid (Spain)

    2004-07-01

    The Boiling Water Reactors (BWR) are susceptible of uncertainties of power when they are operated to low flows of coolant (W) and high powers (P), being presented this situation mainly in the start-up process. The start-up process could be made but sure if the operator knew the value of the stability index Decay reason (Dr) before going up power and therefore to guarantee the stability. The power and the flow are constantly measures, the index Dr could also be considered its value in real time. The index Dr depends on the power, flow and many other values, such as, the distribution of the flow axial and radial neutronic, the temperature of the feeding water, the fraction of holes and other thermohydraulic and nuclear parameters. A simple relationship of Dr is derived leaving of the pattern reduced of March-Leuba, where three independent variables are had that are the power, the flow and a parameter that it contains the rest of the phenomenology, that is to say all the other quantities that affect the value of Dr. This relationship developed work presently and verified its prediction with data of start-up of commercial reactors could be used for the design of a practical procedure practice of start-up, what would support to the operator to prevent this type of events of uncertainty. (Author)

  2. TRAC-BF1/NEM stability methodology for BWR core wide and regional stability analysis

    International Nuclear Information System (INIS)

    A time-series analysis stability methodology is presented based on the TRAC-BF1/NEM coupled code. The methodology presented has a potential application for BWR core-wide and regional stability studies allowed by the 3D capabilities of the code. The stability analysis is performed at two different levels: using the TRAC-BF1 point kinetics model and employing the three-dimensional neutronic transient capability of the NEM code. Point kinetics calculations show power fluctuations when white noise is applied to the inlet mass flow rate of each of the channel components. These fluctuations contain information about the system stability, and are subsequently studied with time-series analysis methods. The analysis performed showed that the reactor core has a low-frequency resonance typical of BWRs. Analysis of preliminary three-dimensional calculations indicates that the power fluctuations do not contain the typical resonance at low frequency. This fact may be related to the limitation of the thermal-hydraulic (T-H) feedback representation through the use of two-dimensional tables for the cross-sections needed for 3D kinetics calculations. The results suggest that a more accurate table look-up should be used, which includes a three-dimensional representation of the feedback parameters (namely, average fuel temperature, average moderator temperature, and void fraction of the T-H cell of interest). Further research is being conducted on improving the cross-section modeling methodology, used to feed the neutron kinetics code for both steady state and transient cases. Also a comprehensive analysis of the code transient solution is being conducted to investigate the nature of the weak dependence of the power response on T-H variations during the performed 3D stability transient calculations

  3. SIMULATE-3K simulation of the Ringhals 1 BWR stability measurements

    International Nuclear Information System (INIS)

    SIMULATE-3K is the transient analysis version of the SIMULATE-3 advanced nodal reactor analysis code. The transient form of the 3-D QPANDA nodal neutronics model has been coupled to a 3-D channel thermal-hydraulics model and a 1-D transient excore peripheral systems model. This paper presents comparisons of SIMULATE-3K calculations and measured BWR stability data from Ringhals Unit 1, Cycles 14-17, as formulated by the OECD/NEACRP. (author)

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

    International Nuclear Information System (INIS)

    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)

  5. 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.

  6. Development of Sirius facility that simulates void-reactivity feedback, and regional and core-wide stability estimation of natural circulation BWR

    International Nuclear Information System (INIS)

    The SIRIUS facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of fluid conditions, power distributions, and fuel rod thermal conductivity time constants, including the normal operating conditions of a typical natural circulation BWR. The results showed that there is a sufficiently wide stability margin under normal operating conditions, even when void-reactivity feedback is taken into account. (author)

  7. Determination of BWR stability characteristics from numerically obtained system responses

    International Nuclear Information System (INIS)

    To determine the stability characteristics of a Boiling Water Reactor with reactor safety codes, one usually uses directly the numerically estimated time responses to a square-shaped or triangular-shaped input perturbation. In this way, the determined stability characteristics depend on the perturbation chosen. The stability characteristics would not depend on the perturbation when they are based on the impulse response. Therefore, a rather straightforward method has been used in order to determine the reactor transfer function and the impulse response from the time signals. The calculated response does not depend on the input perturbation shape used when three things are taken into account. First, the amplitude of the perturbation should not be too large in order to avoid the influence of non-linear effects. Second, the frequency content of the input perturbation should be such that the relevant frequency range for the reactor transfer function is covered. Third, the amplitude and width of the perturbations should be such that the perturbation impact is large enough. For a reference case, the transfer function and impulse response have been determined with RELAP5 to demonstrate the method. With this method one uniquely defines the response for which the stability characteristics are to be evaluated. Furthermore, a direct comparison of impulse response or reactor transfer function between calculation and experiment is possible. (author)

  8. Application of TRACE/PARCS to BWR stability analysis

    International Nuclear Information System (INIS)

    The work described here is the validation of TRACE/PARCS for Boiling Water Reactor stability analysis. A stability methodology was previously developed, verified, and validated using data from the OECD Ringhals stability benchmark. The work performed here describes the application of TRACE/PARCS to all the stability test points from Cycle 14 of the Ringhals benchmark. The benchmark points from Cycle 14 were performed using a half-core symmetric, 325 channel TRACE model. Several parametric studies are performed on test point 10 of Cycle 14. Two temporal difference methods, Semi- Implicit method (SI) and Stability Enhanced Two Step (SETS) method are applied to three different mesh sizes in heated channels with series of time step sizes. The results show that the SI method has a smaller numerical damping than the SETS method. When applying the SI method with adjusted mesh and Courant time step sizes (the largest time step size under the Courant limit), the numerical damping is minimized, and the predicted Decay Ratio (DR) agrees well with the reference values which were obtained from the measured noise signal. The SI method with adjusted mesh and Courant time step size is then applied to all test points of Cycle 14 with 3 types of initiating perturbations, control rod (CR), pressure perturbation, and noise simulation (NS). There is good agreement between the Decay ratios and frequencies predicted by TRACE/PARCS and those from the plant measurements. Sensitivities were also performed to investigate the impact on the decay ratio and natural frequency of the heat conductivity of the gap between fuel and clad, as well as the impact of the pressure loss coefficient of spacers. (authors)

  9. Application of TRACE/PARCS to BWR stability analysis

    International Nuclear Information System (INIS)

    The work described here is the validation of TRACE/PARCS for Boiling Water Reactor stability analysis. A stability methodology was previously developed, verified, and validated using data from the OECD Ringhals stability benchmark. The work performed here describes the application of TRACE/PARCS to all the stability test points from cycle 14 of the Ringhals benchmark. The benchmark points from cycle 14 were performed using a half-core symmetric, 325 channel TRACE model. Several parametric studies are performed on test point 10 of cycle 14. Two temporal difference methods, Semi-Implicit method (SI) and Stability Enhanced Two Step (SETS) method are applied to three different mesh sizes in heated channels with series of time step sizes. The results show that the SI method has a smaller numerical damping than the SETS method. When applying the SI method with adjusted mesh and Courant time step sizes (the largest time step size under the Courant limit), the numerical damping is minimized, and the predicted Decay Ratio (DR) agrees well with the reference values which were obtained from the measured noise signal. The SI method with adjusted mesh and Courant time step size is then applied to all test points of cycle 14 with three types of initiating perturbations, control rod (CR), pressure perturbation, and noise simulation (NS). There is good agreement between the decay ratios and frequencies predicted by TRACE/PARCS and those from the plant measurements. Sensitivities were also performed to investigate the impact on the decay ratio and natural frequency of the heat conductivity of the gap between fuel and clad, as well as the impact of the pressure loss coefficient of spacers

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

    International Nuclear Information System (INIS)

    The Step-III Fuel Type B is a new fuel design for high burn-up operation in BWRs in Japan. The fuel design uses a 9x9 - 9 rod bundle to accommodate the high fuel duty of high burn-up operation and a square water-channel to provide enhanced neutron moderation. The objective of this study is to confirm the thermal-hydraulic stability performance of the new fuel design by tests which simulate the parallel channel configuration of the BWR core. The stability testing was performed at the NFI test loop. The test bundle geometry used for the stability test is a 3x3 heater rod bundle which has about 1/8 of the cross section area of the full size 9x9 - 9 rod bundle. Full size heater rods were used to simulate the fuel rods. For parallel channel simulation, a bypass channel with a 6x6 - 8 heater rod bundle was connected in parallel with the 3x3 rod bundle test channel. The stability test results showed typical flow oscillation features which have been described as density wave oscillations. The stationary limit cycle oscillation extended flow amplitudes to several tens of a percent of the nominal value, during which periodic dry-out and re-wetting were observed. The test results were used for verification of a stability analysis code, which demonstrated that the stability performance of the new fuel design has been conservatively predicted. (author)

  11. TRACE/PARCS validation for BWR stability based on OECD/NEA Oskarshamn-2 benchmark

    International Nuclear Information System (INIS)

    On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event, which culminated in diverging power oscillations with decay ratio greater than 1.3. The event was successfully modeled by TRACE/PARCS coupled code system and the details of the modeling and solution are described in the paper. The obtained results show excellent agreement with the plant data, capturing the entire behavior of the transient including onset of instability, growth of oscillation (decay ratio) and the oscillation frequency. The event allows coupled code validation for BWR with a real, challenging stability event, which challenges accuracy of neutron kinetics (NK), thermal-hydraulics (TH) and TH/NK coupling. The success of this work has demonstrated the ability of 3-D coupled code systems to capture the complex behavior of BWR stability events. The problem is released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (author)

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

    International Nuclear Information System (INIS)

    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)

  13. On the Decay Ratio Determination in BWR Stability Analysis by Auto-Correlation Function Techniques

    International Nuclear Information System (INIS)

    A novel auto-correlation function (ACF) method has been investigated for determining the oscillation frequency and the decay ratio in BWR stability analyses. The neutron signals are band-pass filtered to separate the oscillation peak in the power spectral density (PSD) from background. Two linear second-order oscillation models are considered. These models, corrected for signal filtering and including a background term under the peak in the PSD, are then least-squares fitted to the ACF of the previously filtered neutron signal, in order to determine the oscillation frequency and the decay ratio. Our method uses fast Fourier transform techniques with signal segmentation for filtering and ACF estimation. Gliding 'short-term' ACF estimates on a record allow the evaluation of uncertainties. Numerical results are given which have been obtained from neutron data of the recent Forsmark I and Forsmark II NEA benchmark project. Our results are compared with those obtained by other participants in the benchmark project. The present PSI report is an extended version of the publication K. Behringer, D. Hennig 'A novel auto-correlation function method for the determination of the decay ratio in BWR stability studies' (Behringer, Hennig, 2002)

  14. Uncertainty analysis of the OECD/NRC Oskarshamn-2 BWR stability benchmark

    International Nuclear Information System (INIS)

    On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. The event was successfully modeled by the TRACE/PARCS coupled system code, and further uncertainty analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. This paper shows also how an uncertainty method was implemented for the event. Comparing the calculated uncertainty with the measured uncertainty gives confidence in the BWR stability prediction. (author)

  15. Experimental evaluation of an operator decision aid system for BWR power plants

    International Nuclear Information System (INIS)

    An experiment was carried out to evaluate the effectiveness of an integrated operator decision aid (IODA) system for boiling water reactor (BWR) power plants. The IODA was developed to enhance the operating safety, reliability and quality of plants. It assists an operator's monitoring and diagnostic capabilities under adverse plant situations using computers and color CRT display devices. Three functions - a standby systems management system (SSMS), a disturbance analysis system (DAS), and a post-trip operational guidance (PTOG) system - have been developed as aids corresponding to the operator's various roles in the safe operation of a nuclear power plant. These functions were designed with full consideration of the operator's problem-solving process, and they have been integrated into a comprehensive IODA system for abnormal BWR power plant conditions. To ensure the applicability of the test results to real-life situations, emphasis was placed on establishing realistic test conditions - including a fully equipped experimental control room, a full-scope plant simulator, participation of experienced operating crews, and well-designed test transient scenarios. Experimental data were collected by means of computers and audio-visual devices, and were processed and summarized to facilitate analysis. Qualitative analysis of the test results was performed with emphasis on the decision-making process of an operator. The analysis points to the utility of the IODA as an operator aid, especially in diagnosing adverse plant situations and in formulating strategies for countermeasures. (orig.)

  16. Operational experience of human-friendly control and instrumentation systems for BWR nuclear power plants

    International Nuclear Information System (INIS)

    In recent BWR nuclear power plants in Japan, an advanced centralized monitoring and control system PODIA (Plant Operation by Displayed Information and Automation), which incorporates many operator aid functions, has been in operation since 1985. Main functions of the PODIA system as a computerized operator aid system are as follows. CRT displays for plant monitoring. Automatic controls and operation guides for plant operation. Stand-by status monitoring for engineered safety features during normal operation. Surveillance test procedure guides for engineered safety features. Integrated alarm display. The effectiveness of these functions have been proved through test and commercial operation. It has been obtained that operators have preferred PODIA much more than conventional monitoring and control systems

  17. Fabrication and operating history considerations in assessing relative SCC susceptibility of BWR components

    International Nuclear Information System (INIS)

    The spectrum in material, stress and environmental conditions present in BWR components results in a varied intergranular stress corrosion cracking (IGSCC) response. Fabrication techniques can significantly influence susceptibility to cracking. Components fabricated using cold forming processes tend to experience SCC initiation earlier than similar components subjected to annealing heat treatments, with the degree of acceleration related to the severity and extent of cold work. BWR field experience has shown that SCC initiation and growth was significantly accelerated by the presence of cold work in the upper steam dryer support ring components. Similarly, the propensity for IGSCC initiation and growth in Alloy 182 weld joints may be influenced by microfissuring formed during fabrication and related welding. Ultrasonic inspection results show apparent evidence of accelerated IGSCC initiation due to pre-existing microfissures in nozzle welds and butters. IGSCC is also influenced by irradiation and water chemistry conditions on a creviced component. In addition to demonstrated effects of coolant conductivity and accrued fluence on IGSCC behavior, the localized flux level apparently influences the potential for IGSCC in some components. BWR field experience shows evidence that the propensity for IGSCC in creviced and irradiated control blades is dependent on flux level in higher conductivity plants. Operating with higher purity coolant apparently significantly reduces the potential for IGSCC in creviced components exposed to high flux regions

  18. BWR-incorporated coolant recycling device and operation method therefor

    International Nuclear Information System (INIS)

    An induction electromotive/electric generator is operated as an electric generator by remaining current based on signals of instantaneous power stoppage or loss of power of AC bus in a station to convert an inertia energy stored in flywheels to AC power. It is converted to DC power by way of a controlled inverter device, so that a variable voltage variable frequency power source device (VVVF) can output a DC voltage which can supply an AC voltage having a frequency corresponding to the operation speed of an incorporated type underwater induction electromotive recycling pump (RIP). The voltage is supplied to a series circuit on the side of a rectifier output of the VVVF, so that the operation of the RIP can be continued for a predetermined period of time. Then, the reactor-incorporated coolant recycling device can be operated at a predetermined rated flow rate or predetermined flow rate coast down operation can be conducted thereby enabling to suppress thermal damages of fuel rods, reduce the number of RIP systems and reduce the initial investment of facilities for power generation and cost for inspection and maintenance. (N.H.)

  19. BWR type nuclear plant and start-up operation method

    International Nuclear Information System (INIS)

    During start-up operation of a reactor, the concentration of dissolved oxygen, hydrogen peroxide or ions in primary coolants is controlled at a low level. In addition, hydrogen is injected to the primary coolants to form a stable oxide membrane on the surface of nuclear elements in contact with water. Further, the flow rate of the primary coolants to be supplied to a reactor clean-up system is determined to not less than 2.1% of the feedwater flow rate during rated power operation, and the flow rate of the reactor clean-up system is controlled so that the electroconductivity of the primary coolants is not more than 0.15μS/cm at the upstream of the clean-up device of the reactor clean-up system. Then, the oxidation rate is suppressed and the concentration of ions in water intaken to the oxide membrane is controlled by decreasing the oxide membrane forming speed on the surface of a zirconium alloy and controlling the concentration of ions in the primary coolants. The ingredients which form a tetragonal or cubic crystal structure if they are involved into the oxide membrane can be incorporated in the primary coolants during start-up operation by the control of the ion concentration. (N.H.)

  20. Analysis Applied Multivariate to the Studies of Stability in the Reactors BWR

    International Nuclear Information System (INIS)

    Presently work is presented the application of the analysis multivariate in the studies of stability of reactors BWR. For the confirmation of the applicability of the method of Hilbert Huang is used a group of series acquired neutronic during an outburst in the power station nuclear of Cofrentes. The peculiarity of the analyzed data is that they are not stationary and contaminated by the performance of other systems of the plant, for that that when applying the methods traditional autoregressive to these data, is values non realists of the DR In the work the DR is compared obtained by the methodology presented with the true DR and with the one obtained starting from the application of methods autoregressive to the original sign. The conclusion is evident, the value of the DR obtained by the methodology explained in this work is next to the one True DR that the resulting DR of the application of the method AR to the original sign

  1. BWR stability and bifurcation analysis using reduced order models and system codes: Identification of a subcritical Hopf bifurcation using RAMONA

    International Nuclear Information System (INIS)

    The system code RAMONA, as well as a recently developed BWR reduced order model (ROM), are employed for the stability analysis of a specific operational point of the Leibstadt nuclear power plant. This has been done in order to assess the ROM's applicability and limitations in a quantitative manner. In the context of a detailed local bifurcation analysis carried out using RAMONA in the neighbourhood of the chosen Leibstadt operational point, a bridge is built between the ROM and the system code. This has been achieved through interpreting RAMONA solutions on the basis of the physical mechanisms identified in the course of applying the ROM. This leads, for the first time, to the identification of a subcritical Poincare-Andronov-Hopf (PAH) bifurcation using a system code. As a consequence, the possibility of the so-called correspondence hypothesis is suggested to underline the relationship between a stable (unstable) limit cycle solution and the occurrence of a supercritical (subcritical) PAH bifurcation in the modeling of boiling water reactor stability behaviour

  2. RELAP5/PARCSV2.7 qualification for BWR stability simulations. Application to peach bottom NPP

    International Nuclear Information System (INIS)

    To characterize the unstable behavior of the Peach Bottom Unit 2 BWR, a number of perturbation analyses were performed: arrangements with Philadelphia Electric Company (PECo) were made for conducting different series of Low Flow-Stability Tests at Peach Bottom 2, during the first quarter of 1977. The Low Flow Stability Tests intended to measure the reactor core stability margins at the limiting conditions used in design and safety analysis, providing a one-to-one comparison to design calculations. Stability tests were conducted along the low-flow end of the rated power-flow line, and along the power-flow line corresponding to minimum recirculation pump speed. In this work, three dimensional time domain BWR stability analysis were performed on a new analysis point (PTUPV), which is inside the exclusion region with a core mass flow of 4660.1 kg/s (34% of the core rated mass flow) and total reactor power of 1997.8 MW (60.7 of the core rated reactor power), using the coupled code RELAP5-MOD3.3/PARCSv2.7. This point is achieved departing from test point 3 by the control rod movement as it is usual performed in Nuclear Power Plants. For the core, 48 thermalhydraulic channels have been modeled to represent the active part of the core and one channel for all by-passes. The thermalhydraulic-to-neutronic mapping has been made based on the fundamental and first and second harmonics shapes of the reactor power, calculated with the modal code LAMBDA. For the rest of the plant a coarse nodalization has been adopted for limiting the needed computer resources. For the neutronic code, a nodalization with a 3D core mesh composed with 764 axial nodes has been modeled. A large set of cross section data including 435 compositions has been adopted in neutronic input deck. The purpose of this study is to qualify this coupled code against this kind of 3D complex accidents that take place inside the core. The calculated results show that point PTUPV is an unstable point and the obtained

  3. RELAP5-3D Analysis of Pressure Perturbation at the Peach Bottom BWR During Low-Flow Stability Tests

    International Nuclear Information System (INIS)

    Experimental and theoretical studies about the BWR (Boiling Water Reactor) stability have been performed to design a stable core configuration. BWR instabilities can be caused by inter-dependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In the present work, the pressure perturbation is considered in order to study in detail this type of transient. To simulate this event, including the strong feedback effects between core neutronic and reactor thermal-hydraulics, and to verify core behavior and evaluate parameters related to safety, RELAP5-3D code has been used in the analyses. The simulation was performed making use of Peach Bottom-2 BWR data to predict the dynamics of a real reactor during this type of event. Stability tests were conducted in the Peach Bottom 2 BWR, in 1977, and were done along the low-flow end of the rated power-flow line, and along the power-flow line corresponding to minimum recirculation pump speed. The calculated results are herein compared against the available experimental data. (authors)

  4. Applicability of best-estimate analysis TRACE in terms of natural circulation BWR stability

    International Nuclear Information System (INIS)

    As a part of the international CAMP-Program of the US Nuclear Regulatory Commission (USNRC), the best-estimate code TRACE is validated with the stability database of SIRIUS-N Facility at high pressure. The TRACE code analyzed is version 5 patch level 2. The SIRIUS-N facility simulates thermal-hydraulics of the economic simplified BWR (ESBWR). The oscillation period correlates well with bubble transit time through the chimney region regardless of the system pressure, inlet subcooling and heat flux. Numerical results exhibits type-I density wave oscillation characteristics, since core inlet restriction shifts stability boundary toward the higher inlet subcooling, and chimney exit restriction enlarges instability region and oscillation amplitude. Stability maps in reference to the subcooling and heat flux obtained from the TRACE code agrees with those of the experimental data at 1 MPa. As the pressure increases from 2 MPa to 7.2 MPa, numerical results become much stable than the experimental results. This is because that two-phase frictional loss is underestimate, since the natural circulation flow rate of numerical results is higher by approximately 20% than that of experimental results. (author)

  5. BWR core thermal-hydraulic stability in anticipated transients without SCRAM

    International Nuclear Information System (INIS)

    The potential exists for core thermal hydraulic density wave oscillations to occur in the boiling water reactor (BWR) during anticipated transients without scram (ATWS) events. Conditions which may lead to oscillations in ATWS result either from recirculation pump trip (RPT) from appropriate ATWS signals, or from the failure to scram during core oscillations. A program to evaluate these scenarios will assess the adequacy of the system design and emergency operating procedures relative to the reactor system performance. Evaluation of ATWS events with oscillations is extremely complex, and defining the bases for the analyses is critical to the success of the program. Accurate analysis requires detailed modeling of the sequence of events, the plant configuration, and initial conditions

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. Operation and fuel design strategies to minimise degradation of failed BWR fuel

    International Nuclear Information System (INIS)

    Degradation of failed fuel may result in forced shutdown of the reactor to extract the failed fuel. If this occurs during a time when the price of electricity is high, the cost for this forced shutdown may be very costly. The objective of this paper is to point out the impact of fuel design and also operation strategy on the tendency of failed fuel degradation. The following number of items are discussed in the paper: Failure causes: The dominating causes are debris fretting, PCI and crud/water chemistry related defects. It is recommended to adopt the goal, maximum one defect per year per million rods in the core and to achieve the zero-failure goal for PCI. Models for secondary failure development: Two different secondary degradation scenarios can develop, circumferential cracks or breaks and axial cracks. Models for describing the propagation of secondary defects are given and discussed. The secondary degradation tendency can be delayed and minimized by using fuel cladding with improved corrosion resistance such as cladding with large secondary phase particles and high iron content in the liner layer. Also, the spacer design has a large impact on the tendency for transversal break formation. A spacer that catches the debris at the lower part of the fuel assembly will reduce the risk of getting transversal breaks. On the other hand a spacer that catches the debris in the upper part of the fuel assembly will result in a significant risk of developing transversal breaks in low and intermediate burnup fuel. A new model for data analyses - BwrFuelRelease: A new model, BwrFuelRelease, is presented. This model is an efficient tool for analyses of measured off-gas and reactor water data. The model can replace all currently used methods for analyses of fuel failures. By this model it is possible to detect very small defects, to quantify with high precision the amount of Fissile materials on the core surfaces during operation both with non-defected core and during

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

    International Nuclear Information System (INIS)

    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)

  10. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

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

  11. On the stability of oxides formed under BWR conditions: some observations on a stainless steel and nickel-base alloys

    International Nuclear Information System (INIS)

    BWR fuel CRUD on cladding surfaces, solid corrosion products formed on reactor materials, and solid particles circulating in reactor water all consist of oxides (including hydroxides). These oxides may be exposed to rather different local physical and chemical environments. For example, fuel CRUD is formed on high temperature cladding surfaces and subjected to strong neutron radiation and high turbulent flow. Due to the strong radiolysis reactions the oxides may be exposed to highly oxidizing environments. The oxide films formed on the system piping surfaces, on the other hand, may be exposed to a wide range of temperature and impurity level depending on the reactor circuits in which the materials are used. Some piping materials may not be exposed to neutron radiation. Under different exposure conditions stability of the oxides may vary. A study of oxide stability is important because it may explain why some reactor materials are more prone to corrosion-induced damage than the other materials or why plant radioactivity increases under certain water chemistry conditions or when some materials are used. It may also help to understand the benefit of zinc injection, HWC or other advanced water chemistries. Basically, there are two approaches to study oxide stability in reactor systems. One is thermodynamic method. Another is experimental determination of oxide phases that are sampled from reactor systems or from laboratory loop systems under simulated reactor operation conditions. For the former Pourbaix diagrams are often calculated to show thermodynamic stability regions for various chemical species in a range of temperature, species concentration, potential and pH of interest. In doing thermodynamic calculations it is essential that all possible chemical species are included and their thermodynamic data are properly evaluated especially for the high temperature and pressure range. Omitting an important chemical species may lead to a large error in the prediction or

  12. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study; Analisis de estabilidad de un circuito de recirculacion de un reactor del tipo BWR. Estudio teorico

    Energy Technology Data Exchange (ETDEWEB)

    Salinas H, J.G.; Espinosa P, G. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Gonzalez M, V.M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 04000 Mexico D.F. (Mexico)

    2000-07-01

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  13. An evaluation of an operating BWR piping system damping during earthquake by applying auto regressive analysis

    International Nuclear Information System (INIS)

    The observation of the equipment and piping system installed in an operating nuclear power plant in earthquakes is very umportant for evaluating and confirming the adequacy and the safety margin expected in the design stage. By analyzing observed earthquake records, it can be expected to get the valuable data concerning the behavior of those in earthquakes, and extract the information about the aseismatic design parameters for those systems. From these viewpoints, an earthquake observation system was installed in a reactor building in an operating plant. Up to now, the records of three earthquakes were obtained with this system. In this paper, an example of the analysis of earthquake records is shown, and the main purpose of the analysis was the evaluation of the vibration mode, natural frequency and damping factor of this piping system. Prior to the earthquake record analysis, the eigenvalue analysis for this piping system was performed. Auto-regressive analysis was applied to the observed acceleration time history which was obtained with a piping system installed in an operating BWR. The results of earthquake record analysis agreed well with the results of eigenvalue analysis. (Kako, I.)

  14. Assessment and application of the RAMONA three-dimensional transient code to BWR stability

    International Nuclear Information System (INIS)

    The RAMONA time-domain transient code, with multichannel core hydraulics coupled to a three-dimensional neutronics and fuel model, has been applied to stability analysis of several operating BWRs. The methodology of the analysis is being described and the main results, as compared to stability test data, are presented for four European BWRs. Furthermore, the sensitivity to plant and fuel design parameters is discussed

  15. Application of noble metal chemical addition technology to an operating BWR to mitigate IGSCC of reactor internals

    International Nuclear Information System (INIS)

    Hydrogen Water Chemistry (HWC) has been successfully employed to mitigate the IGSCC of BWR internals over the past decade. However, the use of elevated levels of feed water hydrogen in the BWR results in high operating dose rates due to N16 partitioning into the main steam. Recent studies have shown that the presence of noble metals on reactor internal surfaces, by alloying or by various spray techniques could significantly reduce the hydrogen demand necessary to achieve the IGSCC protection potential of -230 mV(SHE) without the operating dose rate increase. A simpler method of applying noble metal on to reactor internals involve the addition of a noble metal compound into reactor water to cause deposition of noble metal from solution onto surfaces. This noble metal chemical addition (NMCA) technology has been successfully used in numerous laboratory tests to produce a ''noble metal like'' surface on three of the major structural materials, Type 304 SS, Inconel 600 and Alloy 182, used in the nuclear industry. The success of this technology has been tested using constant extension rate tensile (CERT) tests, crack growth rate (CGR) tests and electrochemical corrosion potential (ECP) response tests. The NMCA technology has successfully decreased the ECP of surfaces below -230 mVSHE, prevented crack initiation and mitigated crack growth rates in stoichiometric excess hydrogen in simulated boiling water reactor (BWR) environments. The NMCA treatment of surfaces has drastically lowered the hydrogen demand necessary for IGSCC protection of the materials tested, with no identified side effects including no adverse effects on zircaloy fuel cladding materials. This paper describes the performance of the first NMCA treated BWR over a 12 month period. The paper will also describe the application of NMCA technology to internal components of the BWR by employing the reactor coolant water as the medium of transport for depositing noble metal on in-reactor surfaces. The paper will

  16. Experience of MOX-fuel operation in the Gundremmingen BWR plant: Nuclear characteristics and in-core fuel management

    International Nuclear Information System (INIS)

    After 4 years of good experience with MOX-fuel operation in the BWR plants Gundremmingen units B and C the number of inserted MOX-FAs will be increased in the future continuously. Until now all MOX-FAs are in good condition. Furthermore calculations and measurements concerning zero power tests and tip measurements are in good agreement as expected: all results lead to the conclusion that MOX-FAs can be calculated with the same precision as uranium-FAs. (author)

  17. Application of a reduced order model to BWR corewide stability analysis

    International Nuclear Information System (INIS)

    The determination of system stability parameters from power readings is a problem usually solved by time series techniques such as autoregressive modeling. These techniques are capable of determining the system stability, but ignore the physics of the process and focus on the determination of a nth order linear model. A nonlinear reduced order system is used in conjunction with estimation techniques to present a different approach for stability determination. The simulation of the reduced order model shows the importance of the feedback reactivity imposed by the thermal-hydraulics; the dominant contribution to this feedback is provided by the void reactivity, being a function of power, burnup, power distribution, and in general of the operating conditions of the system. The feedback reactivity is estimated from power measurements and used in conjunction with a reduced order model to determine the system stability properties in terms of the decay ratio

  18. Sludge stabilization operability test report

    International Nuclear Information System (INIS)

    Document provides the results of the Operability Test Procedure performed to test the operability of the HC-21C thermal stabilization process for sludge. The OTP assured all equipment functioned properly and established the baseline temperature profile for glovebox HC-21C

  19. 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

  20. Model for the analysis of transitories and stability of a BWR reactor with fuel of thorium

    International Nuclear Information System (INIS)

    In this work it is described the thermo hydraulic and neutronic pattern used to simulate the behavior of a nucleus of thorium-uranium under different conditions of operation. The analysed nucleus was designed with base to assemblies that operate under the cover-seed concept. The pattern was proven to conditions of stationary state and transitory state. Here it is only presented the simulation of the one SCRAM manual and it is compared in the behavior of a nucleus with UO2. Additionally one carries out an analysis of stability taking into account the four corners that define the area of stability of the map flow-power and to conditions of 100% of flow and 100% of power. The module of stability is based on the pattern of Lahey and Podowsky to estimate the drops of pressure during a perturbation. It is concludes that the behavior of this nucleus is not very different to the one shown by the nuclei loaded with the fuel of UO2. (Author)

  1. Thermal-hydraulic stability analysis of a natural circulation based BWR

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) is a light water cooled and heavy water moderated pressure tube type boiling water reactor. The reactor is designed with the twin objective of utilization of abundant thorium resources and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy and proliferation resistance. In AHWR, it is proposed to remove the core heat by natural circulation during start-up, power raising, normal operation, transients and accidental conditions. A methodology has been presented for analysing the stability behaviour of a multi-channel natural circulation system having different channel layouts. The proposed methodology has been applied to Advanced Heavy Water Reactor (AHWR) and the stable zone of operation for the reactor has been presented

  2. Contribution to BWR stability analysis. Part II: Numerical approach using a reduced order model

    International Nuclear Information System (INIS)

    Highlights: • We study the onset of power oscillations using a reduced order model. • We afford formulae for decay ratios and frequencies near the stability boundary. • We found a non-normal operator associated with the dynamics of the regional mode. • We study some consequences of the meeting of non-normality with nonlinearity. • A comparison between experimental data and model predictions is done. - Abstract: Using the reduced order model and its related analysis done in Suárez-Ántola and Flores-Godoy (submitted for publication), we study some aspects of the onset of power oscillations using numerical methods and digital simulations. From the analytical results we illustrate the usefulness of asymptotic methods to describe the change in behavior of the decay ratio and frequency of oscillations near the stability boundary in the reactor’s parameter space. We study through a dynamical simulation a supercritical Hopf bifurcation in the global mode when the effect of the regional mode on the global mode is neglected. We found that the uncoupled and linearized dynamics of the regional mode is closed related with a non-normal operator. Some of the possible consequences of the non-normality are studied using digital techniques reintroducing the effect of the regional mode on the global mode. A comparison between experimental data and predictions obtained from the present reduced order model is presented

  3. BWR-X experimental reactor for operating in remote sites (thermal-hydraulic study)

    International Nuclear Information System (INIS)

    The primary objective of this indigenous nuclear plant is to power supply no ease accessible places needed of support, like scientific bases and government facilities. What's more, design and built this particular type BWR at low pressure and temperature will give substantial background on nuclear engineering. The paper is intended to treat the nuclear reactor as a heat source in a power cycle and to study the processes by which energy is removed and converted to useful power. (author)

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

    OpenAIRE

    2008-01-01

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

  5. Simulation of the operational monitoring of a BWR with Simulate-3

    International Nuclear Information System (INIS)

    This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

  6. Simulation of the operational monitoring of a BWR with Simulate-3; Simulacion del seguimiento operacional de un reactor BWR con Simulate-3

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez F, J. O.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: ace.jo.cu@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

  7. Evaluation of aging embrittlement of low-carbon austenitic stainless steel weld metal near the BWR operating temperature

    International Nuclear Information System (INIS)

    To evaluate the rate of thermal embrittlement of type 316 low-carbon stainless steel weld metal at BWR operating temperature, aging test at 310degC, 335degC, and 400degC was carried out. Hardness of each ferrite and austenite phase of specimens was measured selectively by microhardness tester. Hardness of ferrite was increased apparently with all three aging temperatures, whereas hardness of austenite was not changed. Changes in the microstructure of ferrite aged at 310degC for 11,000h and 18,000h were analyzed by TEM. It has been revealed that hardening of ferrite with aging at 310degC was result of spinodal decomposition. Arrhenius plots for hardening rate of ferrite at each three aging temperature was plotted, and apparent activation energy was estimated. Using the apparent activation energy, hardening rate at 288degC was discussed. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Spatial and model-order based reactor signal analysis methodology for BWR core stability evaluation

    International Nuclear Information System (INIS)

    A new methodology for the boiling water reactor core stability evaluation from measured noise signals has been recently developed and adopted at the Paul Scherrer Institut (PSI). This methodology consists in a general reactor noise analysis where as much as possible information recorded during the tests is investigated prior to determining core representative stability parameters, i.e. the decay ratio (DR) and the resonance frequency, along with an associated estimate of the uncertainty range. A central part in this approach is that the evaluation of the core stability parameters is performed not only for a few but for ALL recorded neutron flux signals, allowing thereby the assessment of signal-related uncertainties. In addition, for each signal, three different model-order optimization methods are systematically employed to take into account the sensitivity upon the model-order. The current methodology is then applied to the evaluation of the core stability measurements performed at the Leibstadt NPP, Switzerland, during cycles 10, 13 and 19. The results show that as the core becomes very stable, the method-related uncertainty becomes the major contributor to the overall uncertainty range while for intermediate DR values, the signal-related uncertainty becomes dominant. However, as the core stability deteriorates, the method-related and signal-related spreads have similar contributions to the overall uncertainty, and both are found to be small. The PSI methodology identifies the origin of the different contributions to the uncertainty. Furthermore, in order to assess the results obtained with the current methodology, a comparative study is for completeness carried out with respect to results from previously developed and applied procedures. The results show a good agreement between the current method and the other methods

  11. Development of advanced BWR

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

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

  13. Establishment and application of Kuosheng BWR/6 channel flow stability map with RETRANO2/MOD5 code

    International Nuclear Information System (INIS)

    The primary objective of this paper is to determine the channel flow stability map using the RETRAN02/MOD5 code under natural circulation conditions for the Kuosheng Nuclear Power Plant (KNPP). Meanwhile, two transient loci of a KNPP recirculation pump trip (RPT) with decreasing feedwater enthalpy are also drawn on the stability map to assess whether such power oscillation phenomena as the LaSalle-2 event may occur at KNPP. To avoid numerical oscillation of flow stability in time domain analysis, several sensitivity studies are also carried out. The results indicate that both transient loci have sufficient stability margins to unstable boundaries and reveal that KNPP's operations will be stable under such RPT's with decreasing feedwater enthalpy transients. (author)

  14. Stability of (p, Y)-Operator Frames

    Institute of Scientific and Technical Information of China (English)

    Zhi Hua GUO; Huai Xin CAO; Jun Cheng YIN

    2011-01-01

    In this paper we study the stability of (p, Y)-operator frames. We firstly discuss the relations between p-Bessel sequences (or p-frames) and (p, Y)-operator Bessel sequences (or (p, Y)-operator frames). Through defining a new union, we prove that adding some elements to a given (p, Y)-operator frame, the resulted sequence will be still a (p, Y)-operator frame. We obtain a necessary and sufficient condition for a sequence of compound operators to be a (p, Y)operator frame. Lastly, we show that (p, Y)-operator frames for X are stable under some small perturbations.

  15. 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)

  16. Stability performance of the SBWR

    International Nuclear Information System (INIS)

    The stability of boiling water reactors (BWRs) has been a consideration of designers, operators, and regulators since the start-up of early BWRs. Recently, instability phenomena of two different modes (corewide oscillations and regional oscillations) were observed in several BWR plants. These raised additional concerns related to BWR stability. For the simplified BWR (SBWR) design, the approach is to eliminate instability as a concern by assuring a large margin to the stability limits. A conservative set of stability criteria (core decay ratio ≤ 0.4 and channel decay ratio ≤ 0.3 for all expected steady-state operating conditions and moderate frequency transients) is imposed on the SBWR design. During the initial SBWR design, engineering judgment based on operating experience was relied on to select plant parameters that have significant effects on reactor stability. A comparison between the SBWR design and an operating BWR, which had stability test data available, to show that the SBWR design is thermal-hydraulically stable was performed to illustrate this process. After the design was selected, stability analysis was performed to confirm whether the stability criteria had been met. Several iterations were required to obtain the optimal SBWR design. This paper presents a qualitative comparison between the SBWR design and an operating BWR, and the SBWR stability performance as analyzed by a frequency domain code and a three-dimensional thermal-hydraulic code, TRACG

  17. Modernising Sweden's oldest BWR

    International Nuclear Information System (INIS)

    A new range of digital, programmable process control systems has been used to upgrade the instrumentation and control room of Oskarshamn 1's power station. It will enable Sweden's oldest BWR reactor to comply with higher safety levels, improving the operator's overview of production, and reducing the risk of human error. The ''Advant Power'' range of systems will also be used for future planned improvements to the control room. (UK)

  18. Concept of the core for a small-to-medium-sized BWR that does not use control rods during normal operation

    International Nuclear Information System (INIS)

    A small-to-medium-sized boiling water reactor (BWR) with a natural circulation system is being developed for countries where initial investment funds for construction are limited and electricity transmission networks have not been fully constructed. To lighten operators' work load, a core that does not use control rods during normal operation (control rod-free core) was developed by using a neutronics calculation system coupled with core flow evaluation. The control rod-free core had large core power fluctuation with conventional burnable poison design. The target of core power fluctuation was set to less than 10% and was achieved by optimization of burnable poison arrangement. (author)

  19. Crack growth rates for Ni--base alloys with the application to an operating BWR

    International Nuclear Information System (INIS)

    To perform adequate safety assessments of primary components in operating BWRs Crack Growth Rates (CGR) for Stress Corrosion Cracking in Normal Water Chemistry (NWC) as well as Hydrogen Water Chemistry (HWC) are needed. The data behind NUREG 0313 rev 2 was based on laboratory testing of sensitized stainless steels in oxygenated water. This so called NUREG-line overestimates CGRs for operating BWRs with respect to todays specification for water chemistry. In order to suggest new CGRs for Ni-base-alloys in the span from NWC to HWV we performed a literature review. Alloy 600 and welding alloys 182 and 82 were included in the search. The environments were NWC, 'partial' HWC and HWC

  20. Effect of a Sulphate Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 1)

    International Nuclear Information System (INIS)

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. Within WP 3 of this project, the Paul Scherrer Institut (PSI) investigates the effect of water chemistry transients on the EAC crack growth behaviour under periodical partial unloading (PPU) conditions. The present report is a summary of the first PSI test of WP 3 with a Na2SO4 transient. In the first part of the report, the theoretical background on crack growth mechanisms, crack chemistry, mass transport and water chemistry transients as well as a brief literature survey on other water chemistry transient investigations is given. Furthermore, the experimental equipment and test procedure is presented, followed by a summary of the results of PSI test 1 of WP 3. Finally the results are discussed in detail and compared to literature data. In the first part of the experiment, an actively growing EAC crack was generated by PPU in oxygenated high-temperature, high-purity water (T = 288 oC, DO = 8 ppm, SO42- 42- = 368 ppb) of the applied sulphate transient conservatively covered all sulphate transients, which might occur in BWR/normal water chemistry (NWC) practice. After the transient, outlet conductivity was lowered from ca. 1 μS/cm to less than 0.15 μS/cm within 2.6 h by a 'two-loop technique'. No accelerating effect of the sulphate transient on the EAC crack growth of both tested fracture mechanics specimens under highly oxidising BWR/NWC conditions was observed, making it impossible to deterrnine incubation or delay times. The EAC crack growth rates (CGR) before, during and after the sulphate transient were in the range of the 'low-sulphur SCC CGR' of the GE-model. This observation was attributed to 'high-sulphur' crack chemistry conditions, which might

  1. Effect of a Sulphate Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 1)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H. P

    2002-03-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. Within WP 3 of this project, the Paul Scherrer Institut (PSI) investigates the effect of water chemistry transients on the EAC crack growth behaviour under periodical partial unloading (PPU) conditions. The present report is a summary of the first PSI test of WP 3 with a Na{sub 2}SO{sub 4} transient. In the first part of the report, the theoretical background on crack growth mechanisms, crack chemistry, mass transport and water chemistry transients as well as a brief literature survey on other water chemistry transient investigations is given. Furthermore, the experimental equipment and test procedure is presented, followed by a summary of the results of PSI test 1 of WP 3. Finally the results are discussed in detail and compared to literature data. In the first part of the experiment, an actively growing EAC crack was generated by PPU in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm, SO{sub 4}{sup 2-} < 0.6 ppb). Then a sulphate transient was applied. The duration ({approx} 300 h) and the amount of sulphate (SO{sub 4}{sup 2-} = 368 ppb) of the applied sulphate transient conservatively covered all sulphate transients, which might occur in BWR/normal water chemistry (NWC) practice. After the transient, outlet conductivity was lowered from ca. 1 {mu}S/cm to less than 0.15 {mu}S/cm within 2.6 h by a 'two-loop technique'. No accelerating effect of the sulphate transient on the EAC crack growth of both tested fracture mechanics specimens under highly oxidising BWR/NWC conditions was observed, making it impossible to deterrnine incubation or delay times. The EAC crack growth rates (CGR) before, during and after the

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

    International Nuclear Information System (INIS)

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

  3. Achieving stabilization in interferometric logic operations

    Science.gov (United States)

    Zavalin, Andrey I.; Shamir, Joseph; Vikram, Chandra S.; Caulfield, H. John

    2006-01-01

    Interferometric systems with amplitude beam splitters can implement reversible operations that, on detection, become Boolean operators. Being passive, they consume no energy, do not limit the operating bandwidth, and have negligible latency. Unfortunately, conventional interferometric systems are notoriously sensitive to uncontrolled disturbances. Here the use of polarization in a common-path interferometric logic gate with and without polarization beam splitters is explored as an attractive alternative to overcome those difficulties. Two of three device configurations considered offer significant stability and lower drive modulator voltage as advantages over the previous systems. The first experimental tests of such a system are reported. Common-path interferometry lends itself to even more stability and robustness by compatibility with no-air-gap, solid optics.

  4. OPERATING STABILITY OF MINERAL WOOL PRODUCTS

    Directory of Open Access Journals (Sweden)

    Perfilov Vladimir Aleksandrovich

    2016-03-01

    Full Text Available Creating an effective insulation envelope of the building is possible only using high-quality materials, preserving their characteristics both in the early stages of operation, and for the whole billing period. It is an important opportunity to assess the thermal insulation properties and predict its changes over time directly in the conditions of the construction site. The products based on mineral fibers (rock and glass wool, basalt fiber are the most widely used type of insulating materials in the domestic construction. Therefore, the operational stability valuation methods must be primarily created for this group of products. The methodology for assessing the thermal insulation properties includes two main components: testing equipment and methodology for assessing the operational stability. The authors tested the methodology of the accelerated testing and prediction of durability for mineral wool products of laminated, corrugated and volume-oriented structures. The test results give good convergence with the methods recommended by the building regulations. Application of thermal insulation materials are an effective way to form the thermal envelope of the building, reducing energy costs and increasing the durability of building structures. The material properties are determined by their structure, which is formed during the technological impacts.

  5. Application of a lattice filter algorithm to BWR noise monitoring and diagnosis

    International Nuclear Information System (INIS)

    a method for BWR noise monitoring and diagnosis based on an adaptive lattice filter is presented in this paper. It is specifically intended for applying to on-line real time BWR stability monitoring during reactor operation relatively close to unstable regimes, where fast tracking of any stability state change is essential. The authors present a set of algorithms for estimating stability parameters, i.e. decay ratio and peak frequency, for detecting anomaly occurrences based on the pattern recognition methodology, and for calculating signal DC component and mean amplitude. A unique feature of the method is the unified approach for the signal processing, in that these algorithms are all executed based on model identification using the adaptive lattice filter. The present method was applied to BWR noise data collected during stability tests at Forsmark 1 and 2 in Sweden. The result demonstrates that the present method is capable of fast tracking of stability state changes. The paper stresses that the evaluation of the process state by combining the results from the stability parameters estimation, pattern comparison, and checking the behavior of DC component and mean amplitude of process signals is important for obtaining reliable estimate and adequate information on the stability state, especially when the reactor is operated near instability

  6. A novel auto-correlation function method and FORTRAN codes for the determination of the decay ratio in BWR stability analysis

    International Nuclear Information System (INIS)

    A novel auto-correlation function (ACF) method has been investigated for determining the oscillation frequency and the decay ratio in BWR stability analyses. The report describes not only the method but also documents comprehensively the used and developed FORTRAN codes. The neutron signals are band-pass filtered to separate the oscillation peak in the power spectral density (PSD) from background. Two linear second-order oscillation models are considered. The ACF of each model, corrected for signal filtering and with the inclusion of a background term under the peak in the PSD, is then least-squares fitted to the ACF estimated on the previously filtered neutron signals, in order to determine the oscillation frequency and the decay ratio. The procedures of filtering and ACF estimation use fast Fourier transform techniques with signal segmentation. Gliding 'short-time' ACF estimates along a signal record allow the evaluation of uncertainties. Some numerical results are given which have been obtained from neutron signal data offered by the recent Forsmark I and Forsmark II NEA benchmark project. They are compared with those from other benchmark participants using different other analysis methods. (author)

  7. A novel auto-correlation function method and FORTRAN codes for the determination of the decay ratio in BWR stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Behringer, K

    2001-08-01

    A novel auto-correlation function (ACF) method has been investigated for determining the oscillation frequency and the decay ratio in BWR stability analyses. The report describes not only the method but also documents comprehensively the used and developed FORTRAN codes. The neutron signals are band-pass filtered to separate the oscillation peak in the power spectral density (PSD) from background. Two linear second-order oscillation models are considered. The ACF of each model, corrected for signal filtering and with the inclusion of a background term under the peak in the PSD, is then least-squares fitted to the ACF estimated on the previously filtered neutron signals, in order to determine the oscillation frequency and the decay ratio. The procedures of filtering and ACF estimation use fast Fourier transform techniques with signal segmentation. Gliding 'short-time' ACF estimates along a signal record allow the evaluation of uncertainties. Some numerical results are given which have been obtained from neutron signal data offered by the recent Forsmark I and Forsmark II NEA benchmark project. They are compared with those from other benchmark participants using different other analysis methods. (author)

  8. BWR fuel performance

    International Nuclear Information System (INIS)

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

  9. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

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

  10. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    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 B4C 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)

  11. Model for the analysis of transitories and stability of a BWR reactor with fuel of thorium; Modelo para el analisis de transitorios y de estabilidad de un reactor BWR con combustible de torio

    Energy Technology Data Exchange (ETDEWEB)

    Nunez C, A. [CNSNS, 03020 Mexico D.F. (Mexico)]. E-mail: anunezc@cnsns.gob.mx; Espinosa P, G. [UAM-I, 09340 Mexico D.F. (Mexico); Francois L, J.L. [Fac. de Ingenieria, UNAM 62550 Jiutepec, Morelos (Mexico)

    2004-07-01

    In this work it is described the thermo hydraulic and neutronic pattern used to simulate the behavior of a nucleus of thorium-uranium under different conditions of operation. The analysed nucleus was designed with base to assemblies that operate under the cover-seed concept. The pattern was proven to conditions of stationary state and transitory state. Here it is only presented the simulation of the one SCRAM manual and it is compared in the behavior of a nucleus with UO{sub 2}. Additionally one carries out an analysis of stability taking into account the four corners that define the area of stability of the map flow-power and to conditions of 100% of flow and 100% of power. The module of stability is based on the pattern of Lahey and Podowsky to estimate the drops of pressure during a perturbation. It is concludes that the behavior of this nucleus is not very different to the one shown by the nuclei loaded with the fuel of UO{sub 2}. (Author)

  12. BWR type nuclear power plant and operation method therefor and method of forming oxide membrane on the surface of the constitutional member in contact with water

    International Nuclear Information System (INIS)

    In a BWR type nuclear power plant, an oxide membrane is formed on the surface of the constitutional members of a reactor primary system to be in contact with water while keeping the reactor water at a pH of 7.5 or less based on a room temperature and keeping a temperature of reactor water at 250degC or higher for 250 hours or more and then adding alkaline water to control the pH within a range of from 7.5 to 9.0 based on the room temperature and keeping the reactor water temperature to 250degC or higher for 100 hours or more. This process is conducted during the reactor shut down state and during the operation period from the time of the reactor shut down state to the time of the rated power operation state of the electric power generator. Then, a corrosion resistant oxide membrane with less involvement of radioactive ions can be formed, thereby enabling to improve corrosion resistance of nuclear fuel elements and suppressing the dose rate on the surface of pipelines of a primary coolant system, accordingly, operator's radiation dose rate can be reduced upon periodical inspection. (N.H.)

  13. Results of VGB research work with respect to operation of BWR pipes made of austenitic SS; Ergebnisse des VGB-Forschungsvorhabens zur Absicherung des Betriebsverhaltens austenitischer Staehle in SWR-Rohrleitungen

    Energy Technology Data Exchange (ETDEWEB)

    Kilian, R. [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Bruemmer, G. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany)

    1998-11-01

    The VGB research project was to examine and characterize various, operation-induced impacts on the crack formation in stabilized austenitic steels, caused by intercrystalline stress corrosion cracking as a result of sensitization after chromium depletion at the grain boundaries. The results of this project as well as available operating experience show that the measures taken so far for the future operation of the German BWR plants, for avoiding in these plants intercrystalline stress corrosion cracking, correspond to the state of the art and achieve the wanted purpose. These measures are: use of optimized material W-No. 1.4550 with reduced carbon contents; use of optimized welding techniques for reducing the heat input and the welding shrinkage (cold deformation.); optimized preparation of welding work in order to avoid shape defects during welding (eg. edge misalignment, defective mash welds); reduction of tensile stresses occurring during welding; compliance with the recent VGB water chemistry code. (orig./CB) [Deutsch] Das VGB-Forschungsvorhaben sollte verschieden gelagerte Einfluesse auf die Rissbildung im Betrieb von stabilisierten austenitischen Staehlen, verursacht durch interkristalline Spannungsrisskorrosion infolge Sensbilisierung durch Chromverarmung an den Korngrenzen, systematisch erfassen. Aus den Forschungsergebnissen dieses VGB-Programms sowie den bisher vorliegenden Betriebserfahrungen ist festzuhalten, dass die bisher durchgefuehrten Massnahmen fuer den zukuenftigen Betrieb der deutschen SWR-Anlagen zur Vermeidung von interkristalliner Spannungsrisskorrosion zielgerichtet waren und dem heutigen Wissensstand entsprechen. Diese Massnahmen sind: 1. Einsatz von optimiertem Werkstoff W.-Nr. 1.4550 mit abgesenktem Kohlenstoffgehalt; 2. Einsatz von optimierten Schweissverfahren zur Verminderung der Waermeeinbringung und zur Verringerung des Schweissschrumpfes (Kaltverformung.); 3. Durchfuehrung einer optimierten Schweissnahtvorbereitung zur Vermeidung

  14. BWR radiation control: plant demonstration

    International Nuclear Information System (INIS)

    The first year's progress is presented for a four-year program intended to implement and evaluate BRAC radiation reduction operational guidelines at the Vermont Yankee BWR and to document the results in sufficient detail to provide guidance to other BWR owners. Past operational, chemistry and radiation level data have been reviewed to provide a historical base of reference. Extensive sampling and chemistry monitoring systems have been installed to evaluate plant chemistry status and the effects of program implemented changes. Radiation surveys and piping gamma scans are being performed at targeted locations to quantify radiation level trends and to identify and quantify piping isotopics. Contact radiation levels on the recirculation line at Vermont Yankee have been increasing at a rate of 175 mR/h-EFPY since 1978. A materials survey of feedwater and reactor components in contact with the process liquid has been performed to identify sources of corrosion product release, particularly cobalt and nickel. A feedwater oxygen injection system has been installed to evaluate the effects of oxygen control on feedwater materials corrosion product releases. A baseline performance evaluation of the condensate treatment and reactor water cleanup systems has been completed. Data on organics and ionics at Vermont Yankee have been obtained. A methodology of BWR feedwater system layup during extended outages was developed, and an evaluation performed of layup and startup practices utilized at Vermont Yankee during the fall 1980 and 1981 refueling outages

  15. Effect of a Chloride Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 2)

    International Nuclear Information System (INIS)

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the second test of working package (WP) 3 with a NaCl transient, performed at Paul Scherrer Institut (PSI). In the first part of the experiment, an actively growing EAC crack with a crack growth rate (CGR) in the range of the 'low-sulphur SCC line' of the GE-model was generated by periodical partial unloading (PPU) in oxygenated high-temperature, high-purity water (T = 288 oC, DO = 8 ppm). Then a chloride transient of 49 ppb Cl- was applied for ∼40 h. After this transient, the load level was reduced and the loading conditions were changed to pure cyclic loading. Thereupon a second transient with a chloride concentration of 49 ppb was applied. In both RPV steels, the first chloride transient of 49 ppb Cl- resulted in an acceleration of the EAC crack growth by more than one order of magnitude and in fast, stationary SCC crack growth during the constant load phase of the PPU cycles at KI values 1/2. 3 h after adding chloride to the high-purity water, the EAC CGR started to increase in the high-sulphur RPV steel during the constant load phase of a PPU cycle and after 20 h a stationary EAC CGR value in the range of the 'high-sulphur SCC curve' of the GE-model was reached. After 5 h in high-purity water, the crack growth began to slow down after a partial unloading cycle and 15 h later it reached again a stationary CGR value in the range of the 'low-sulphur SCC curve' of the GE-model. The second chloride transient did not result in an acceleration of the crack growth in both investigated specimens. This was explained by crack closure effects, which occurred in both specimens after the reduction of the load. The CGR during

  16. 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.

  17. Comparison study of the thermal mechanical performance of fuel rods during BWR fuel preconditioning operations using the computer codes FUELSIM and FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Ortiz V, J.; Castillo D, R., E-mail: rafael.pantoja10@yahoo.com.m [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The safety of nuclear power plants requires monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behaviour under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. In the operation of a nuclear power reactor, pre-conditioning simulations are necessary to determine in advance limit values for the power that can be generated in a fuel rod during any power ramp, and mainly during reactor startup, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR is performed. This study includes two types of fuel rods: one from a fuel assembly design with array 8 x 8, and the other one from a 10 x 10 fuel assembly design, and a comparison of the thermal-mechanical performance between the two different rod designs is performed. The performance simulations were performed by the code FUELSIM, and compared against results previously obtained from similar simulation with the code FEMAXI-V. (Author)

  18. Multiattribute prediction of terrain stability above underground mining operations

    OpenAIRE

    Vujić Slobodan; Miljanović Igor; Milutinović Aleksandar; Đorđević Dragan; Gojković Nebojša; Gajić Grozdana

    2011-01-01

    This paper is dedicated to the problem of stability prediction of the terrain above underground mining operations. After the initial introduction to the problem, then the short analysis of the model approaches used to solve it, and giving the algorithm for rock massif stability prediction, we describe the concept of the multiattirbute terrain stability prediction method. The application of the multiattribute prediction method for stability of the terrain above underground mining operati...

  19. BWR 90 - the advanced BWR of the 1990s

    International Nuclear Information System (INIS)

    The future global role of nuclear power will be determined by its ability to provide economical and safe energy. Nuclear power, like any other substantial contributor to the world's energy needs, must be generated at an acceptable cost and with negligible environmental effects. Besides, it must achieve and maintain a socially reasonable level of public acceptance, which in turn is not necessarily governed by rational assessments of the true safety and environmental impact of nuclear power. The ABB Atom approach to this situation can best be characterized as a 'cautious evolution'; for the next decade the company will largely base its offerings to the market on its 'evolutionary' light water reactor design, the BWR 90. This design builds closely on the experience from successful construction and operation of its predecessor, the BWR 75 design. In 1995 and 1996, plants of this design achieved an average load factor greater than the 87% set by EUR; the two BWR units at Olkiluoto in Finland are among the very best performing plants in the world, with an average load factor of 94% over the last 7 years. The continued LWR design development focuses on meeting requirements from utilities as well as new regulatory requirements. A particular emphasis is put on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimization of buildings and containment design to decrease construction time and costs, and selection of materials as well as maintenance and operating procedures to even further reduce occupational radiation exposures. (orig.)

  20. The boiling water reactor BWR 90

    International Nuclear Information System (INIS)

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

  1. Cutting BWR feedwater crud levels further

    International Nuclear Information System (INIS)

    Reducing iron input to the BWR primary system is an important first step in decreasing radiation fields and occupational exposure. For feedwater iron, the specified optimum concentration is 0.1-0.5 ppb, as demonstrated by the newest ''low crud'' plants operating in Japan. Recent advances in condensate filtration will achieve the levels needed for optimised water chemistry and promise great benefits. Of particular interest is a newly developed filterdemineraliser septum that separates the filtration and ion exchange functions to allow each to be specifically optimised for BWR conditions. (author)

  2. A BWR fuel channel tracking system

    International Nuclear Information System (INIS)

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

  3. Development and operational experiences of an automated remote inspection system for interior of primary containment vessel of a BWR

    International Nuclear Information System (INIS)

    A prototype was developed for an automated remote inspection system featuring continuous monitoring of the working status of major components inside the primary containment vessel of a boiling water reactor. This inspection system consists of four units, or vehicles, which are towed by a trolley chain along a monorail; a complex coaxial cable for data transmission and for power supply; and an operator's console. A TV camera, microphone, thermometer, hygrometer, and ionization chamber are mounted on the various units. After several months' testing under high-ambient temperature, the system was installed in the Tokai-2 power station of Japan Atomic Power Company for in situ tests

  4. Procedures for using expert judgment to estimate human-error probabilities in nuclear power plant operations. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Seaver, D.A.; Stillwell, W.G.

    1983-03-01

    This report describes and evaluates several procedures for using expert judgment to estimate human-error probabilities (HEPs) in nuclear power plant operations. These HEPs are currently needed for several purposes, particularly for probabilistic risk assessments. Data do not exist for estimating these HEPs, so expert judgment can provide these estimates in a timely manner. Five judgmental procedures are described here: paired comparisons, ranking and rating, direct numerical estimation, indirect numerical estimation and multiattribute utility measurement. These procedures are evaluated in terms of several criteria: quality of judgments, difficulty of data collection, empirical support, acceptability, theoretical justification, and data processing. Situational constraints such as the number of experts available, the number of HEPs to be estimated, the time available, the location of the experts, and the resources available are discussed in regard to their implications for selecting a procedure for use.

  5. Local stabilizer codes in 3D without string logical operators

    CERN Document Server

    Haah, Jeongwan

    2011-01-01

    We suggest concrete models for self-correcting quantum memory by reporting examples of local stabilizer codes in 3D that have no string logical operators. Previously known local stabilizer codes in 3D all have string-like logical operators, which make the codes non-self-correcting. We introduce an algebraic definition of "logical string segments" to avoid difficulties in defining one dimensional objects in discrete lattices. We prove that every string-like logical operator of our code can be deformed to a disjoint union of short segments, and each segment is in the stabilizer group. The code has surface-like logical operators whose partial implementation has unsatisfied stabilizers along its boundary.

  6. OPERATING STABILITY OF MINERAL WOOL PRODUCTS

    OpenAIRE

    Perfilov Vladimir Aleksandrovich; Pilipenko Anton Sergeevich; Pyataev Evgeniy Ravil’evich

    2016-01-01

    Creating an effective insulation envelope of the building is possible only using high-quality materials, preserving their characteristics both in the early stages of operation, and for the whole billing period. It is an important opportunity to assess the thermal insulation properties and predict its changes over time directly in the conditions of the construction site. The products based on mineral fibers (rock and glass wool, basalt fiber) are the most widely used type of insulating materia...

  7. MAAP BWR application guidelines

    International Nuclear Information System (INIS)

    The MAAP Thermal-Hydraulic Qualification and Application Project has as its objective to identify those thermal-hydraulic phenomena modeled in MAAP which are important in predicting severe accident sequences, to qualify those models and to provide guidelines for use of the code. This report provides user guidelines for use of the BWR version of MAAP. The report includes a discussion of the important features of the BWR that are modeled in MAAP, the MAAP modeling of phenomena important to predicting severe accidents and user guidelines for several accident sequences

  8. Establishing operational stability--developing human infrastructure.

    Science.gov (United States)

    Gomez, Max A; Byers, Ernest J; Stingley, Preston; Sheridan, Robert M; Hirsch, Joshua A

    2010-12-01

    Over the past year, Toyota has come under harsh scrutiny as a result of several recalls. These well publicized mishaps have not only done damage to Toyota's otherwise sterling reputation for quality but have also called into question the assertions from a phalanx of followers that Toyota's production system (generically referred to as TPS or Lean) is the best method by which to structure one's systems of operation. In this article, we discuss how Toyota, faced with the pressure to grow its business, did not appropriately cadence this growth with the continued development and maintenance of the process capabilities (vis a vis the development of human infrastructure) needed to adequately support that growth. We draw parallels between the pressure Toyota faced to grow its business and the pressure neurointerventional practices face to grow theirs, and offer a methodology to support that growth without sacrificing quality. PMID:21990652

  9. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    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)

  10. Common Cause Failure Analysis of Control Rods and Drives in the Swedish and Finnish BWR Plants. Operating Experiences in 1983 - 2003

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, Tuomas [Avaplan Oy, Espoo (Finland)

    2006-11-15

    The control rod and drives in a Boiling Water Reactor (BWR) constitute a highly redundant system. The reliability of the system is determined by how well the design withstands dependencies, as Common Cause Failures (CCFs). This report upgrades an earlier data collection on CCFs of control rod and drives (SKI Report 1996:77) to more recent years, with the objective to report the data to ICDE project (International Common Cause Failure Data Exchange) and to the safety analysts in the Nordic countries. The operating experiences were analyzed at the BWRs of former Asea-Atom design, comprising 9 units in Sweden and Olkiluoto 1 and 2 in Finland, covering years 1983 - 2003. A new logical scheme was developed to classify interconnected failure modes of the two redundant functions for reactivity shutdown, fast hydraulic insertion and slower screw insertion of control rods. The scheme makes an explicit distinction between the different attributes of the failure event: - affected function - affected movement direction - detectability - criticality, i.e. inoperable control rod function versus only degraded functionality Another novel idea emerged for grouping the events according to generic failure mechanism. The generic classes will help to organize and structure the information efficiently, because in most cases within a class, the failure modes prove to be same, or there are only a few alternatives to chose from. From the set of 72 candidate cases, altogether 27 actual or more significant potential CCFs were screened out. Special emphasis was placed to identify any multiple failure or degradation indicating that adjacent rods would be more vulnerable to failure, because such phenomena are far more critical for reactivity shutdown as compared to failure of randomly placed rods. Only slight tendency of position dependence could be determined. Another positive insight is that the events, where foreign objects caused the jamming of rod insertion, were separated by both

  11. Common Cause Failure Analysis of Control Rods and Drives in the Swedish and Finnish BWR Plants. Operating Experiences in 1983 - 2003

    International Nuclear Information System (INIS)

    The control rod and drives in a Boiling Water Reactor (BWR) constitute a highly redundant system. The reliability of the system is determined by how well the design withstands dependencies, as Common Cause Failures (CCFs). This report upgrades an earlier data collection on CCFs of control rod and drives (SKI Report 1996:77) to more recent years, with the objective to report the data to ICDE project (International Common Cause Failure Data Exchange) and to the safety analysts in the Nordic countries. The operating experiences were analyzed at the BWRs of former Asea-Atom design, comprising 9 units in Sweden and Olkiluoto 1 and 2 in Finland, covering years 1983 - 2003. A new logical scheme was developed to classify interconnected failure modes of the two redundant functions for reactivity shutdown, fast hydraulic insertion and slower screw insertion of control rods. The scheme makes an explicit distinction between the different attributes of the failure event: - affected function - affected movement direction - detectability - criticality, i.e. inoperable control rod function versus only degraded functionality Another novel idea emerged for grouping the events according to generic failure mechanism. The generic classes will help to organize and structure the information efficiently, because in most cases within a class, the failure modes prove to be same, or there are only a few alternatives to chose from. From the set of 72 candidate cases, altogether 27 actual or more significant potential CCFs were screened out. Special emphasis was placed to identify any multiple failure or degradation indicating that adjacent rods would be more vulnerable to failure, because such phenomena are far more critical for reactivity shutdown as compared to failure of randomly placed rods. Only slight tendency of position dependence could be determined. Another positive insight is that the events, where foreign objects caused the jamming of rod insertion, were separated by both

  12. BWR internal cracking issues

    International Nuclear Information System (INIS)

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

  13. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

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

  14. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

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

  15. The design of large natural circulation BWR's

    International Nuclear Information System (INIS)

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

  16. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

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

  17. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

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

  18. BWR AXIAL PROFILE

    Energy Technology Data Exchange (ETDEWEB)

    J. Huffer

    2004-09-28

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

  19. 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

  20. THEORETICAL CONSIDERATIONS REGARDING THE AUTOMATIC FISCAL STABILIZERS OPERATING MECHANISM

    Directory of Open Access Journals (Sweden)

    Gondor Mihaela

    2012-07-01

    Full Text Available This paper examines the role of Automatic Fiscal Stabilizers (AFS for stabilizing the cyclical fluctuations of macroeconomic output as an alternative to discretionary fiscal policy, admitting its huge potential of being an anti crisis solution. The objectives of the study are the identification of the general features of the concept of automatic fiscal stabilizers and the logical assessment of them from economic perspectives. Based on the literature in the field, this paper points out the disadvantages of fiscal discretionary policy and argue the need of using Automatic Fiscal Stabilizers in order to provide a faster decision making process, shielded from political interference, and reduced uncertainty for households and business environment. The paper conclude about the need of using fiscal policy for smoothing the economic cycle, but in a way which includes among its features transparency, responsibility and clear operating mechanisms. Based on the research results the present paper assumes that pro-cyclicality reduces de effectiveness of the Automatic Fiscal Stabilizer and as a result concludes that it is very important to avoid the pro-cyclicality in fiscal rule design. Moreover, by committing in advance to specific fiscal policy action contingent on economic developments, uncertainty about the fiscal policy framework during a recession should be reduced. Being based on logical analysis and not focused on empirical, contextualized one, the paper presents some features of AFS operating mechanism and also identifies and systematizes the factors which provide its importance and national individuality. Reaching common understanding on the Automatic Fiscal Stabilizer concept as a institutional device for smoothing the gap of the economic cycles across different countries, particularly for the European Union Member States, will facilitate efforts to coordinate fiscal policy responses during a crisis, especially in the context of the fiscal

  1. Reliability innovations for AREVA NP BWR fuel

    International Nuclear Information System (INIS)

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

  2. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

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

  3. Service experience of BWR pressure vessels

    International Nuclear Information System (INIS)

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

  4. The road not taken: addressing corruption during stability operations

    OpenAIRE

    Revell, Brian; Nemeth, Ryan-Ross

    2015-01-01

    Approved for public release; distribution is unlimited The United States has spent the last 14 years engaging in combat operations in Iraq and Afghanistan that have aimed, in part, to rebuild two dysfunctional states. However, after billions of dollars in development money, thousands of soldiers’ lives lost, and over a decade of time, neither of these countries has achieved the desired degree of stability; both states remain fragile and sources of regional and global insecurity. This thesi...

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

    International Nuclear Information System (INIS)

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

  6. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    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)

  7. Effect of Loading Transients on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Tests 3 and 4)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.-P

    2003-04-01

    Within the CASTOC-project (5{sup t}h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the third and fourth test of working package (WP) 3 with loading transients, performed at Paul Scherrer Institut (PSI). Two different low-alloy steels (20 MnMoNi 5 5, 0.015 wt.% S and 22 NiMoCr 3 7, 0.007 wt. %S) were investigated in oxygenated high-temperature, high-purity water (T = 240 {sup o}C, DO = 400 ppb) in a daisy chain at two different load ratios (R = 0.8 and 0.2). In the first part of the experiments, asymmetrical saw tooth loading with different rise times {delta}t{sub R} of the load and different loading frequencies were applied. Then the loading conditions were changed to an asymmetrical trapezoid waveform loading (periodical partial unloading, PPU) and the hold time {delta}t{sub H} at maximum load was varied. In the final phase of WP 3 PSI tests 3 and 4 the SCC behaviour was investigated under constant load. With decreasing loading frequency the corrosion fatigue (CF) crack advance per cycle {delta}a/{delta}N{sub EAC} of material A increased. Sustained EAC crack growth could be maintained down to low frequencies of 10{sup -5} Hz. The time-based crack growth rate (CGR) da/dt{sub EAC} decreased with decreasing frequency. In material B no effect of the loading frequency could be resolved. Up to a hold time of 1 h at maximum constant load the CGR da/dt{sub EAC} seemed to be independent of the hold time. Above hold times of 1 h the CGR decreased and dropped down to CGR values in the range or below the BWR VIP 60 SCC disposition lines. This behaviour was observed in both investigated materials. The cycle-based CGR {delta}a/{delta}N{sub EAC} remained approximately constant with increasing hold time. The

  8. Effect of Loading Transients on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Tests 3 and 4)

    International Nuclear Information System (INIS)

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the third and fourth test of working package (WP) 3 with loading transients, performed at Paul Scherrer Institut (PSI). Two different low-alloy steels (20 MnMoNi 5 5, 0.015 wt.% S and 22 NiMoCr 3 7, 0.007 wt. %S) were investigated in oxygenated high-temperature, high-purity water (T = 240 oC, DO = 400 ppb) in a daisy chain at two different load ratios (R = 0.8 and 0.2). In the first part of the experiments, asymmetrical saw tooth loading with different rise times ΔtR of the load and different loading frequencies were applied. Then the loading conditions were changed to an asymmetrical trapezoid waveform loading (periodical partial unloading, PPU) and the hold time ΔtH at maximum load was varied. In the final phase of WP 3 PSI tests 3 and 4 the SCC behaviour was investigated under constant load. With decreasing loading frequency the corrosion fatigue (CF) crack advance per cycle Δa/ΔNEAC of material A increased. Sustained EAC crack growth could be maintained down to low frequencies of 10-5 Hz. The time-based crack growth rate (CGR) da/dtEAC decreased with decreasing frequency. In material B no effect of the loading frequency could be resolved. Up to a hold time of 1 h at maximum constant load the CGR da/dtEAC seemed to be independent of the hold time. Above hold times of 1 h the CGR decreased and dropped down to CGR values in the range or below the BWR VIP 60 SCC disposition lines. This behaviour was observed in both investigated materials. The cycle-based CGR Δa/ΔNEAC remained approximately constant with increasing hold time. The low-frequency corrosion fatigue (LFCF) CGR Δa/ΔNEAC of material A and B

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

    International Nuclear Information System (INIS)

    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)

  10. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

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

  11. Stability of klystron operation as a function of input parameters

    International Nuclear Information System (INIS)

    Instabilities in the operation of a high-power multicavity klystron have been studied as a function of various input parameters. Among these, the focusing magnetic field, rf input power and tuning of the second harmonic cavity were observed to have a striking effect in deciding the regions of unstable operation. One region of instability could be identified with a value of magnetic field corresponding to cyclotron resonance, at the operating frequency. The magnetic field value in the region of the input end, and the second harmonic cavity, was found to have a more decisive influence on stability, than the magnetic field at the output end. A hysteresis effect has been associated with the instabilities. This could be explained in terms of multipactor phenomena. These observations indicate, among other things, that in designing new tubes, it might be desirable that the operating magnetic field stay well away from cyclotron resonance conditions, especially in the gun region, and the cavities at the input end. The experimental techniques can be used for optimizing the operating conditions of existing tubes

  12. 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)

  13. Entanglement-assisted operator codeword stabilized quantum codes

    Science.gov (United States)

    Shin, Jeonghwan; Heo, Jun; Brun, Todd A.

    2016-05-01

    In this paper, we introduce a unified framework to construct entanglement-assisted quantum error-correcting codes (QECCs), including additive and nonadditive codes, based on the codeword stabilized (CWS) framework on subsystems. The CWS framework is a scheme to construct QECCs, including both additive and nonadditive codes, and gives a method to construct a QECC from a classical error-correcting code in standard form. Entangled pairs of qubits (ebits) can be used to improve capacity of quantum error correction. In addition, it gives a method to overcome the dual-containing constraint. Operator quantum error correction (OQEC) gives a general framework to construct QECCs. We construct OQEC codes with ebits based on the CWS framework. This new scheme, entanglement-assisted operator codeword stabilized (EAOCWS) quantum codes, is the most general framework we know of to construct both additive and nonadditive codes from classical error-correcting codes. We describe the formalism of our scheme, demonstrate the construction with examples, and give several EAOCWS codes

  14. The armenian power system operation stability investigation accounting putting new power systems into operation

    International Nuclear Information System (INIS)

    The description of the power systems operation stability failure caused by the system significant emergency states occurred during the last working period in Armenian and USA power systems is performed. With the use of PSSTME-31 software portfolio of Siemens Firm a design model is developed and transient electromechanical process calculations for Armenian power system are performed. The accuracy of the model is checked by comparing real-time transient state parameters and their reproduction calculation results.The Armenia - Iran current power transmission lines permissible limit under the condition of the static and dynamic stability requirements and in case of the new thermal power units maintenance are defined

  15. Nonuniform Exponential Stability and Instability of Evolution Operators in Banach Space

    OpenAIRE

    Mihaela Tomescu; Andrea Minda

    2006-01-01

    In this paper is presenting a parallel between nonuniform exponential stability and nonuniform exponential instability of evolution operators in Banach spaces, beginning to present the concept of the evolution operator with nonuniform exponential decay, respectively growth, next with the concept of the nonuniform stability, respectively instability, nonuniform exponential stability, respectively instability, nonuniform integrable stability, respectively instability and...

  16. Nonuniform Exponential Stability and Instability of Evolution Operators in Banach Space

    Directory of Open Access Journals (Sweden)

    Mihaela Tomescu

    2006-10-01

    Full Text Available In this paper is presenting a parallel between nonuniform exponential stability and nonuniform exponential instability of evolution operators in Banach spaces, beginning to present the concept of the evolution operator with nonuniform exponential decay, respectively growth, next with the concept of the nonuniform stability, respectively instability, nonuniform exponential stability, respectively instability, nonuniform integrable stability, respectively instability and relationship between this concepts.

  17. Thermal-hydraulics in BWR

    International Nuclear Information System (INIS)

    In the heat transferring flow in BWRs, the heightening of heat transfer performance accompanying the development of new fuel for the purpose of reducing spent fuel generation and the improvement of fuel economy, the heightening of performance and the reduction of size of various heat exchangers, the development of the safety devices, of which the constitution is simple, the reliability is high, and the operation is easy, and so on are expected. As for ABWRs, thermal output is 3926 MW, and electricity output is 1356 MW. The system constitution of ABWR is shown. The main change from BWR to ABWR is the adoption of internal pumps, reinforced concrete containment vessels and electric control rod drive. For evaluating the limit output of high burnup fuel assemblies, the subchannel analysis and the effect that spacers exert to the limit output are explained. The heat transferring flow in moisture separation heater, condenser and feed water heater is reported. The heat transferring flow in passive containment vessel cooling system of water wall type and condensing type is described. (K.I.)

  18. Natural convection type BWR reactor

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    A denatured (U-233/Th)O2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O2-fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O2-fueled BWR should perform similar to a UO2-fueled BWR under all operating conditions. A (Pu/Th)O2-fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO2-fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

  2. Development of ECP models for BWR applications

    International Nuclear Information System (INIS)

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

  3. 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)

  4. Study on the feasibility of 1300 MWe class simplified BWR plant

    International Nuclear Information System (INIS)

    A range of power levels for 1000 MWe-1500 MWe natural circulation core was found to be feasible from the thermal hydraulic performance standpoint by our sensitivity analysis. In this study, we selected a power level of 1300 MWe that is expected to satisfy Japanese Utilities needs. After we set the RPV configuration, we will study the detailed comprehensive analysis so that we can confirm the technical feasibility of large scaled simplified BWR. RPV inner diameter 7.5 m, which can be manufactured with current technology and present facilities, and the chimney height of 8.5 m was selected. After a preliminary design of the core and fuel was carried out, the natural circulation core flow was calculated by EASHAP code. The stability evaluation during normal operation is analyzed and a major transient analysis is conducted. The design of the core and fuel is evaluated based on PANACEA code. The detailed analysis shows that a 1300 MWe class natural circulation core satisfies the thermal and stability criteria. The containment system, which consists of the drywell and suppression chamber, is determined with supporting containment pressure-temperature analytical response. The layout inside the primary containment vessel that is applicable to a RPV incorporating the 1300 MWe core is approximately arranged. From the above, it is confirmed that 1000 MWe is not technical upper power limit of the simplified BWR plant. (author)

  5. BwrCrud: Development and validation of a new code for improved simulation of activity transport in BWR primary systems

    International Nuclear Information System (INIS)

    During the nineties, much reactor experience and significant new findings have been acquired. Thus, it became necessary to improve the previous code for computer model. Development of an improved code (named as BwrCrud) for modeling of activated corrosion products in BWR systems is progressing to model the influence of factors such as iron and zinc flow, HWC operation and fuel failures in a better way. The present version of the Code, BwrCrud 1.0 was tested in the Kashiwazaki Kariwa 5 (KK-5) plant and it was concluded that the cord can work satisfactorily and manage to model the KK-5 behavior appropriately in consideration of some uncertainties in input and verification data. Phase 2 of the project is to be completed in the beginning of 1999. (M.N.)

  6. High Pressure Boiling Water Reactor HP-BWR

    International Nuclear Information System (INIS)

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

  7. BWR type reactor

    International Nuclear Information System (INIS)

    In a coolant circulation in BWR type reactors, since the mixed stream of steam fluid undergoes a great resistance, the pressure loss due to the flow rate distribution when the coolants flow from the upper plenum into the stand pipe is increased upon passing stand pipe. Also in the spontaneous recycling reactor, pressure loss is still left upon passing the swirling blade of a gas-liquid separator. In view of the above, a plurality of vertical members each having a lower end opened to a gas-liquid two phase boundary and an upper end directly suspended from a steam dryer to the gas-liquid separator. The liquid droplets from the 2-phase boundary heated in the reactor core and formed into a mixed gas-liquid 2-phase stream is directed in the vertical direction accompanied with the steam. The liquid droplets spontaneously fallen by gravity from greater ones successively and the droplets in the steam abutted against the vertical member are fallen as a liquid membrane. Thus, the gas-liquid separation is conducted, the dry steam is directly flown into the steam dryer, thereby capable of providing a gas-liquid separator having gas-liquid separation performance with lower loss than usual. (N.H.)

  8. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  9. State of the art of second international exercise on benchmarks in BWR reactors

    International Nuclear Information System (INIS)

    This is a second in series of Benchmarks based on data from operating Swedish BWRs. The first one concerned measurements made in cycles 14,15 16 and 17 at Ringhals 1 Nuclear Power Plant and addressed predictive power of analytical tools used in BWR stability analysis. Part of the data was disclosed only after participants had provided their results. This work has been published in the report: NEA/NSC/DOC(96)22, November 1996. In this report it was recognised that there is a need for better qualification of the applied noise analysis methods. A follow up Benchmark was thus proposed dedicated to the analysis of time series data and including the evaluation of both global and regional stability of Forsmarks 1 and 2 Nuclear Power Plant. In this second Benchmark have participated Forsmarks Kraftgrupp AB,NEA Nuclear Science Committee, CSN Consejo de Seguridad Nuclear and Department of Chemical and Nuclear Engineering of Polytechnic University of Valencia. (Author)

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

    International Nuclear Information System (INIS)

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

  11. Experience and reliability of Framatome ANP's PWR and BWR fuel

    International Nuclear Information System (INIS)

    Based on three decades of fuel supply to 169 PWR and BWR plants on four continents, Framatome ANP has a very large database from operating experience feedback. The performance of Framatome PWR and BWR fuel is discussed for the period 1992-2001 with special emphasis on fuel failures, countermeasures and their effectiveness. While PWR fuel performance in most reactors has been good, the performance in some years did suffer from special circumstances that caused grid-to-rod fretting failures in few PWRs. After solving this problem, fuel of all types showed high reliability again. Especially the current PWR fuel products AFA 3G, HTP, Mark B and Mark BW showed a very good operating performance. Fuel reliability of Framatome ANP BWR fuel has been excellent over the last decade with average annual fuel rod failure rates under 1x10-5 since 1991. More than 40% of all BWR fuel failures in the 1992-2001 decade were caused by debris fretting. The debris problem has been remedied with the FUELGUARDTM lower tie plate, and by reactor operators' efforts to control the sources of debris. PCI, the main failure mechanism in former periods, affected only 10 rods. All of these rods had non-liner cladding. (author)

  12. Design of a redundant meteorological station for a BWR reactor

    International Nuclear Information System (INIS)

    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)

  13. 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)

  14. ATRIUMTM Fuel - Continuous Upgrading for High Duty BWR Plants

    International Nuclear Information System (INIS)

    AREVA NP is a supplier of nuclear fuel assemblies and associated core components to Boiling Water Reactors worldwide, representing today more than 60 000 fuel assemblies. Since first delivered in 1992, ATRIUMTM10 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. Among them, the latest versions are ATRIUMTM 10XP and ATRIUMTM 10XM fuel assemblies which have been delivered to several utilities worldwide. During six years of operation experience reaching a maximum fuel assembly burnup of 66 MWd/kgU, no fuel failure of ATRIUMTM 10XP/XM occurred. Regular upgrading of the fuel assemblies' reliability and performance has been made possible thanks to AREVA NP's continuous improvement process and the 'Zero tolerance for failure' program. In this frame, the in-core behavior follow-up, manufacturing experience feedback and customer expectations are the bases for setting improvement management objectives. As an example, most fuel rod failures observed in the past years resulted from debris fretting and Pellet Cladding Interaction (PCI) generally caused by Missing Pellet Surface. To address these issues, the development of the Improved FUELGUARDTM debris filter was initiated and completed while implementation of chamfered pellets and Cr doped fuel will address PCI aspects. In the case of fuel channel bow issue, efforts to ensure dimensional stability at high burnup levels and under challenging corrosion environments have been done resulting in material recommendations and process developments. All the described solutions will strongly support the INPO goal of 'Zero fuel failures by 2010'. In a longer perspective, the significant trend in nuclear fuel operation is to increase further the discharge burnup and/or to increase the reactor power output. In the majority of nuclear power plants worldwide, strong efforts in power up-rating were made and are still ongoing. Most

  15. Nuclear transmutation characteristics of reduced moderation BWR (Thesis)

    International Nuclear Information System (INIS)

    In the present thesis, the nuclear transmutation characteristics of reduced moderation BWR, which decides the spent fuel characteristics and its safety in its nuclear fuel cycle, were investigated and compared with other types of reactors. The major conclusions were obtained as follows: The decay heat and radioactivity from FPs increases in fuel burn-up. However, they which normalized with burn-up are small for the reactor with low specific power and long operation period due to the decay during the long operation period. Breeder type of reduced moderation BWR shows low decay heat and radioactivity from FPs because of the long operation period approximately 3000 days which realized by the high conversion ratio. That also shows low decay heat and radioactivity from actinide nuclides due to the hard spectrum. MA recycling reactor of high conversion type of reduced moderation BWR was designed. The neptunium, which has large impact for environmental burden from the viewpoint of nuclide transport analysis, can be incinerated approximately 40% of loaded inventory which corresponds to 22 units of LWR per year. LLFP (99Tc, 129I, 135Cs) transmutation by breeder type of reduced moderation BWR was estimated. As a result, the support factor cannot be lower than unity for each LLFP nuclides. In other words, the reduced moderation BWR cannot reduce LLFP because the LLFP target cannot be loaded inner of the reactor core due to the small margin of core specification. It is expected that these results and the characteristics of other types of reactor shown in the present study benefit the discussion for various nuclear fuel cycle options. (author)

  16. Dynamic analysis of BWR scram reactivity characteristics

    International Nuclear Information System (INIS)

    An extensive study of BWR scram reactivity behavior is presented. It is based on a space-time analysis of a BWR/4 code using the two-dimensional (R, Z) dynamics code BNL-TWIGL which includes a two-phase thermal-hydraulic model. Calculations were made of the sensitivity of scram to physical quantities such as initial control rod position and power distribution, scram speed, system pressure and varying inlet flow rate and temperature. The end-of-cycle Haling operating condition was found to give rise to the limiting scram reactivity function. Even with scram a power surge was found to be possible with severely decreasing inlet temperature. Calculations were also made to find the effect on scram of commonly used modeling approximations. These included the effect of neglecting delayed neutrons (conservative), using a time invariant void distribution (non-conservative) and defining point kinetics parameters such as reactivity, amplitude function and generation time in terms of different weighting functions. The importance of defining point kinetics parameters consistent with their use in plant transient analyses was demonstrated with particular emphasis on the role of ''residual reactivity''

  17. 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.

  18. Operative stabilization of the cervical spine in cases of metastases

    International Nuclear Information System (INIS)

    Because of intraspinal space requirement or instability of the cervical spine, metastases may lead to a compression of the spinal cord and/or the nerve roots. Early decompression and stabilization prevents unavoidable tetraparesis. In most cases quality of life can be maintained. The indication and methodology of cervical spine stabilization as well as the results of treatment in 18 patients are demonstrated. (orig.)

  19. Contingency programs for BWR pipe cracking

    International Nuclear Information System (INIS)

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

  20. Transmutation of minor actinide using thorium fueled BWR core

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Stability and monotonicity of Lotka-Volterra type operators

    CERN Document Server

    Mukhamedov, Farrukh

    2009-01-01

    In the present paper, we study Lotka-Volterra (LV) type operators defined in finite dimensional simplex. We prove that any LV type operator is a surjection of the simplex. After, we introduce a new class of LV-type operators, called $M$LV type. We prove convergence of their trajectories and study certain its properties. Moreover, we show that such kind of operators have totaly different behavior than ${\\mathbf{f}}$-monotone LV type operators.

  3. Diagnosis of nonlinear BWR oscillations using TRAC/BF1

    International Nuclear Information System (INIS)

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

  4. Real time simulation of the main steam system of a BWR nuclear power station

    International Nuclear Information System (INIS)

    This paper presents a real time model of the main steam system for a BWR 675 MW power plant unit. The model includes the start up and shut down of the system, where the steam flow is very small or non existent and phenomena like condensation can occur, changing drastically the effects observed from those of normal operation at medium or high loads. Severe transients are also contemplated. Consistency and stability tests were done to the model, and it was validated for steady state using plant design data. During transients the model's results were compared with the predictions of the Final Safety Analysis Report (FSAR) for the prototype unit, and it was found that the model's response follow the expected trends

  5. 14 CFR 29.672 - Stability augmentation, automatic, and power-operated systems.

    Science.gov (United States)

    2010-01-01

    ... functioning of stability augmentation or other automatic or power-operated system is necessary to show compliance with the flight characteristics requirements of this part, the system must comply with § 29.671 of... stability augmentation system or in any other automatic or power-operated system which could result in...

  6. 14 CFR 27.672 - Stability augmentation, automatic, and power-operated systems.

    Science.gov (United States)

    2010-01-01

    ... stability augmentation system or in any other automatic or power-operated system which could result in an... systems. (b) The design of the stability augmentation system or of any other automatic or power-operated system must allow initial counteraction of failures without requiring exceptional pilot skill or...

  7. Swedes repair BWR thermal fatigue cracks

    International Nuclear Information System (INIS)

    In connection with an accident at the Barsebaeck-2 NPP the causes of cracking in steel pipe fittings of the BWR type reactor cooling system are investigated. In the course of testing carried out by the methods of gamma radiography and liquid penetrant inspection the cracks 10-100 mm long and with depth up to 10 mm are found. The most of the cracks is concentrated in regions near pipe fittings in the direction of water stream flow. The cause of crack formation is the thermal stress arising during mixing the water with different temperatures in particular, the feedwater having at normal operational conditions temperature of 180 deg C and the emergency cooling system water with the temperature of 270 deg C. The conclusion is drawn on the necessity of designing the new configurations of joints which are able to withstand the temperature gradients

  8. Radiation source term reduction in BWR plants

    International Nuclear Information System (INIS)

    This series of slides presents: the collective radiation exposures at US and European BWRs; the European experience with source term reduction measures (normal water chemistry - NWC): zinc addition, stellite replacement, full system decontamination; the effects of evolving water chemistries/US experience. The conclusions are summarized as follows: worldwide reduction of collective radiation exposures at BWRs by following the ALARA principle; zinc addition proven option for source term reduction for NWC and hydrogen water chemistry (HWC) plants; reducing feedwater iron has been proven to reduce dose rates - as operational observations in the US indicate; optimized feedwater iron is very important for fuel performance under all modes of water chemistry (HWC, Zn, and noble metal chemical addition (NMCA)); minimize 59Co sources/stellite, follow the ALARA principle; full system decontamination (FSD) plus zinc injection is an attractive option for reducing reactor coolant system (RCS) dose rates of mature BWR plants

  9. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

  10. Peach bottom cycle 2 stability analysis using RELAP5/PARCS

    International Nuclear Information System (INIS)

    Boiling channels and systems may oscillate owing to the behaviour of the liquid-steam mixture used for removing the thermal power. A thermal-hydraulic system may be unstable under particular operating conditions. Two kinds of power oscillation have been observed in BWR cores. One is an in-phase (core-wide) and the other is an out-of-phase (regional) oscillation. Since the above feature can make detection more difficult, the latter oscillation is potentially more severe. The problem is well known since the design of the first BWR system. However, to improve the safety systems of these reactors, it is necessary to be able to detect in a reliable way these oscillations from the neutronic signals. The purpose of this work is to characterize the unstable behaviour of a BWR. Within this study, it has been performed a number of perturbation analysis. The coupled codes RELAP5-Mod3.3/PARCS have used for the simulation of the transients. Validation has been performed against Peach Bottom-2 Low-Flow Stability Test PT3. Three dimensional time domain BWR stability analysis were performed on test point 3 for the core wide oscillation mode. In this transient dynamically complex events take place, i.e., neutron kinetics is coupled with thermal-hydraulics and an in-phase oscillation has been developed. The calculated results are compared against the available experimental data. (author)

  11. Robust Stability and H∞ Stabilization of Switched Systems with Time-Varying Delays Using Delta Operator Approach

    Directory of Open Access Journals (Sweden)

    Chen Qin

    2013-01-01

    Full Text Available This paper considers the problems of the robust stability and robust H∞ controller design for time-varying delay switched systems using delta operator approach. Based on the average dwell time approach and delta operator theory, a sufficient condition of the robust exponential stability is presented by choosing an appropriate Lyapunov-Krasovskii functional candidate. Then, a state feedback controller is designed such that the resulting closed-loop system is exponentially stable with a guaranteed H∞ performance. The obtained results are formulated in the form of linear matrix inequalities (LMIs. Finally, a numerical example is provided to explicitly illustrate the feasibility and effectiveness of the proposed method.

  12. Wall stabilized operation in high beta NSTX plasmas

    International Nuclear Information System (INIS)

    The National Spherical Torus Experiment, NSTX, has demonstrated the advantages of low aspect ratio geometry in accessing high βt ≡ 2μ0/B02 and βN ≡ 108t>aB0/Ip. Experiments have reached βt = 39% through boundary and profile optimization and βN = 6.8 utilizing moderate current profile modification. High βN plasmas can exceed the ideal no-wall stability limit, βNno-wall, for periods much greater than the wall eddy current decay time. Resistive wall mode (RWM) physics is studied to understand mode stabilization in these plasmas. The toroidal mode spectrum of unstable RWMs has been measured with mode number n up to 3. The critical rotation frequency of Bondeson-Chu, Ωcrit = ωA/(4q2) describes well the RWM stability of NSTX plasmas when applied over the entire rotation profile and in conjunction with the ideal stability criterion. Rotation damping and global rotation collapse observed in plasmas exceeding βNno-wall contrasts the damping observed during tearing mode activity and can be described by drag due to neoclassical toroidal viscosity (NTV) in the helically perturbed field of an ideal displacement. Resonant field amplification of an applied n = 1 field perturbation has been measured and increases with increasing βN. Equilibria are reconstructed including measured ion and electron pressure, toroidal rotation, and flux iso-surface constraint in plasmas with core rotation ωφ/ωA up to 0.48. Peak pressure shifts of 11% of the minor radius from the magnetic axis have been reconstructed. (author)

  13. Hexagonal eutectic solidification patterns operating near a marginal stability point

    OpenAIRE

    Perrut, Mikael; Akamatsu, Silvère; Bottin-Rousseau, Sabine; Faivre, Gabriel

    2008-01-01

    We study the long-time dynamics of hexagonal directional-solidification patterns in bulk samples of a transparent eutectic alloy using an optical method which permits real-time observation of the growth front. A slow dilatation of the patterns due to a slight curvature of the isotherms drives the system into a permanent regime, close to the threshold for the rod splitting instability. Thus an apparently minor instrumental imperfection suffices to maintain the system near a marginal stability ...

  14. BWR zero pressure containment

    International Nuclear Information System (INIS)

    This patent describes the operation of a nuclear reactor system, the system including a containment defining a drywall space wherein a nuclear reactor is disposed, there being a suppression pool in the containment with the suppression pool having a wetwell space above a level of the pool to which an non-condensable gases entering the suppression pool can vent. It comprises: continuously exhausting the wetwell space to remove gas mixture therefrom while admitting inflow of air from an atmospheric source thereof to the wetwell during normal operation by blocking off the inflow during a loss-of-coolant-accident whenever a pressure in the wetwell space is above a predetermined value, and subjecting the gas subsequent to its removal from the wetwell to a treatment operation to separate any particulate material entrained therein from the gas mixture

  15. Transportation activities for BWR fuels at NFI

    International Nuclear Information System (INIS)

    Nuclear Fuel Industries, LTD. (NFI) supplies fuel assemblies for both PWR and BWR in Japan. We can also manage transportation of the fuel assemblies from our fabrication facilities to the nuclear power plants of Japanese utilities. For the transportation of fuel assembly, we designed and fabricated the transportation containers to meet the requirements of the IAEA regulations, and licensed in Japan. This paper introduced the recent activity and R and D of NFI concerning transportation of BWR fuel assembly. NT-XII transportation container was developed for fresh BWR fuel assemblies. NT-XII container consists of inner container and outer container. Two BWR fuel assemblies with up to 5 wt.-% 235-U enriched are enclosed in an inner container. In the concept of NT-XII container design, we made the best priority to transportation efficiency, as well as ensuring fuel integrity during transportation. NT-XII has been used since 2002 in Japan. Thanks to the lightening weight of containers, the number of containers to be loaded to one transportation truck was increased up to 9 containers (equivalent to 18 fuel assemblies) compared with former type container (NT-IV transportation container) which can be loaded up to 6 containers (equivalent to 12 fuel assemblies). In addition to the design of brand-new container, we promote the improvement of the packaging methods. In Japan, in order to reduce the damage to the fuel rod and fuel spacer while transporting, polyethylene sleeves which are called 'packing separators' are inserted in the rod-to-rod gap of fuel assembly. However, packing separators requires time and cost for the installation at fuel fabrication facility. In the same way, huge time and cost are needed for removal of packing separators at nuclear power plant. For the improvement of preparation efficiency before and after transportation, we investigated the influence of vibration to fuel integrity in case of transportation without packing separators. Based on the above

  16. Actively-stabilized photomultiplier tube base for vacuum operation

    International Nuclear Information System (INIS)

    An actively stabilized photomultiplier tube (PMT) base design for an Amperex XP-2262B PMT is described. Positive-negative-positive transistors are used as low-impedance current sources to maintain constant voltages on the last three dynodes. This technique results in a highly stable, low-power tube base ideal for use with low-duty-factor beams, such as those found at the Clinton P. Anderson Meson Physics Facility. Furthermore, because of the low power usage of this large design, these bases can be sealed in a heat-conductive, electrically insulating material and used in a vacuum

  17. Stabilizing operation point technique based on the tunable distributed feedback laser for interferometric sensors

    Science.gov (United States)

    Mao, Xuefeng; Zhou, Xinlei; Yu, Qingxu

    2016-02-01

    We describe a stabilizing operation point technique based on the tunable Distributed Feedback (DFB) laser for quadrature demodulation of interferometric sensors. By introducing automatic lock quadrature point and wavelength periodically tuning compensation into an interferometric system, the operation point of interferometric system is stabilized when the system suffers various environmental perturbations. To demonstrate the feasibility of this stabilizing operation point technique, experiments have been performed using a tunable-DFB-laser as light source to interrogate an extrinsic Fabry-Perot interferometric vibration sensor and a diaphragm-based acoustic sensor. Experimental results show that good tracing of Q-point was effectively realized.

  18. BANKING SYSTEM STABILITY: COMMERCIAL AND CO-OPERATIVE BANKS

    Directory of Open Access Journals (Sweden)

    Dumitru-Cristian OANEA

    2014-04-01

    Full Text Available Commercial banks and co-operative banks are credit institutions, but there are some differences between the main operations proceeded by each of them. Based on these specific characteristics, we want to identify the manner in which financial crisis affected their activity. As we all know, the financial crisis had a major impact in the United States, the “natal” country of the crisis, because great banks such as Lehman Brothers or Merrill Lynch have bankrupted. Even if the Romanian banking system was not affected by such catastrophic situations, surely the financial crisis had a significant impact on it. This topic is worth to be analysed, because we would be able to identify the risk differences between these two types of business: commercial banks versus co-operative banks.

  19. BWR type reactor

    International Nuclear Information System (INIS)

    No channel box is mounted to a fuel assembly, but a partition plate for separating coolant flow channels between each of fuel bundles is disposed between each of fuel bundles along the direction of height for the reactor core instead of the channel box. The partition plate has a shape surrounding the fuel bundles only in a specific region, or so that coolant flow channels for a plurality of fuel bundles of identical output are integrated. As a result, cross-flow of coolants can be prevent without channel box and, further, radial expansion of the channel box can be eliminated. As the same time, the bending for the entire assembly due to the irradiation growth of the channel box is also eliminated and structural stability can be attained without using upper grid plates. Further, it is possible to minimize the pressure loss caused between the upper and lower portions of the assembly and it is possible to adjsut with respective thermohydrodynamic properties of the high conversion region and the burner region. (K.M.)

  20. BWR radiation control: plant demonstration. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    The first year's progress is presented for a four-year program intended to implement and evaluate BRAC radiation reduction operational guidelines at the Vermont Yankee BWR and to document the results in sufficient detail to provide guidance to other BWR owners. Past operational, chemistry and radiation level data have been reviewed to provide a historical base of reference. Extensive sampling and chemistry monitoring systems have been installed to evaluate plant chemistry status and the effects of program implemented changes. Radiation surveys and piping gamma scans are being performed at targeted locations to quantify radiation level trends and to identify and quantify piping isotopics. Contact radiation levels on the recirculation line at Vermont Yankee have been increasing at a rate of 175 mR/h-EFPY since 1978. A materials survey of feedwater and reactor components in contact with the process liquid has been performed to identify sources of corrosion product release, particularly cobalt and nickel. A feedwater oxygen injection system has been installed to evaluate the effects of oxygen control on feedwater materials corrosion product releases. A baseline performance evaluation of the condensate treatment and reactor water cleanup systems has been completed. Data on organics and ionics at Vermont Yankee have been obtained. A methodology of BWR feedwater system layup during extended outages was developed, and an evaluation performed of layup and startup practices utilized at Vermont Yankee during the fall 1980 and 1981 refueling outages

  1. BWR chromium chemistry

    International Nuclear Information System (INIS)

    This report addresses the concern about higher total specific conductivity in the reactor recirculation loop water due to the chromate ion. This concern is particularly high at plants where all other ionic species have been reduced through careful attention to makeup and condensate polisher operations. An EPRI Chromate Workshop was held in November 1990 to consider the issues raised by observed levels of chromate ion (generally 5 to 50 ppB). While BWRs on normal water chemistry were the only ones observing chromate, even plants on hydrogen water chemistry (HWC) observe sharp spikes of conductivity due to chromate whenever the hydrogen supply was interrupted after a reasonably long HWC operational period. The consensus of the workshop attendees was that chromate was not a concern as an agent causing pipe cracking compared to the more common species such as chloride and sulfate. However, the data are somewhat ambiguous for levels of chromate above 50 ppB. Adjustments to the weighing factors for the various ionic species in the industry chemistry performance index are suggested to allow for the known relative higher aggressiveness of other species relative to that of chromate

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

    International Nuclear Information System (INIS)

    The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  3. 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)

  4. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    programs that GE has participated in and describes the different options and approaches that have been used by various utilities in their design basis programs. Some of these variations deal with the scope and depth of coverage of the information, while others are related to the process (how the work is done). Both of these topics can have a significant effect on the program cost. Some insight into these effects is provided. The final section of the paper presents a set of lessons learned and a recommendation for an optimum approach to a design basis information program. The lessons learned reflect the knowledge that GE has gained by participating in design basis programs with nineteen domestic and international BWR owner/operators. The optimum approach described in this paper is GE's attempt to define a set of information and a work process for a utility/GE NSSS Design Basis Information program that will maximize the cost effectiveness of the program for the utility. (

  5. 14 CFR 23.672 - Stability augmentation and automatic and power-operated systems.

    Science.gov (United States)

    2010-01-01

    ... provided for any failure in the stability augmentation system or in any other automatic or power-operated system that could result in an unsafe condition if the pilot was not aware of the failure. Warning systems must not activate the control system. (b) The design of the stability augmentation system or...

  6. Stability Analysis for Operation of DG Units in Smart Grids

    DEFF Research Database (Denmark)

    Pouresmaeil, Edris; Shaker, Hamid Reza; Mehrasa, Majid;

    2015-01-01

    This paper presents a multifunction control strategy for the stable operation of Distributed Generation (DG) units during grid integration. The proposed control model is based on Direct Lyapunov Control (DLC) theory and provides a stable region for the appropriate operation of DG units during grid...... integration. Using DLC technique in DG technology can provide the continuous injection of maximum active power in fundamental frequency from the DG source to the grid, compensating all reactive power and harmonic current components of grid-connected loads through the integration of DG link into the grid....... Application of this concept can guarantee to reduce the stress on the grid during the energy demand peak. Simulation results are presented to demonstrate the proficiency and performance of the proposed DLC technique in DG technology....

  7. Studies on the transient operation and stability of fast reactors

    International Nuclear Information System (INIS)

    These studies form part of the general programme of perfecting calculation methods for fast reactors. The basic formulae are given for the layouts used, i.e. the classic kinetic and thermal exchange equations, etc. A description is then given of the digital computer methods employed for studying the stable functioning of the reactor and of the methods used for transient operation studies. Finally, some examples of application are discussed and a comparison is made with parallel studies on the same subject. (author)

  8. Winning the war at home: stability operations strategy for homeland security

    OpenAIRE

    Shaughnessy, Joshua

    2016-01-01

    Approved for public release; distribution is unlimited While much progress has been made in American homeland security, there remains a lack ofstrategic consensus for state and local agencies. Communities are increasingly outspoken in their expectations of unified homeland security efforts. Remarkably, the U.S. military has a long history of using stability operations strategy to execute similar missions in foreign and domestic settings. Stability operations strategy is a non-traditional U...

  9. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

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

  10. BWR control rod design using tabu search

    International Nuclear Information System (INIS)

    An optimization system to get control rod patterns (CRP) has been generated. This system is based on the tabu search technique (TS) and the control cell core heuristic rules. The system uses the 3-D simulator code CM-PRESTO and it has as objective function to get a specific axial power profile while satisfying the operational and safety thermal limits. The CRP design system is tested on a fixed fuel loading pattern (LP) to yield a feasible CRP that removes the thermal margin and satisfies the power constraints. Its performance in facilitating a power operation for two different axial power profiles is also demonstrated. Our CRP system is combined with a previous LP optimization system also based on the TS to solve the combined LP-CRP optimization problem. Effectiveness of the combined system is shown, by analyzing an actual BWR operating cycle. The results presented clearly indicate the successful implementation of the combined LP-CRP system and it demonstrates its optimization features

  11. The Delft desire facility for studies on (natural circulation) BWR primary system statics and dynamics

    International Nuclear Information System (INIS)

    A test facility for research on BWR core statics and dynamics was designed and built in Delft. The loop, DESIRE, consists of a BWR fuel assembly, a riser, condenser and a downcorner section. Freon-12 is used as a coolant. Presently, research on this facility is focused on investigations of the physical aspects of natural-circulation cooling and reactor kinetic stability. To this end, an artificial feedback from in-core void fraction to heating power is being established. The void fraction is determined on a sub-channel level by measuring the transmission of a collimated gamma beam

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

    International Nuclear Information System (INIS)

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

  13. Design of an Interpolated Controller for Stabilization of a Plant with Variable Operating Condition

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    In this paper, the stabilization of a linear SISO plant with variable operating condition is considered. The plant is described by a linear interpolation of proper stable co-prime factorizations of the transfer functions at two representative operating points. An interpolation of the stabilizing controllers for the representative models is designed to stabilize the plant, and the necessary and sufficient condition for the plant to be stabilized by the proposed controller is presented using the Nevanlinna-Pick interpolation theory [4, 5]. It is shown that the class of stabilization plants via the proposed controller in the paper is larger than that by the controller in reference [3]. An example is also given to illustrate this fact.

  14. 14 CFR 25.672 - Stability augmentation and automatic and power-operated systems.

    Science.gov (United States)

    2010-01-01

    ... system or in any other automatic or power-operated system which could result in an unsafe condition if...) The design of the stability augmentation system or of any other automatic or power-operated system... exceptional pilot skill or strength, by either the deactivation of the system, or a failed portion thereof,...

  15. Neutron flux and fluence determination for BWR reactors

    International Nuclear Information System (INIS)

    Measurements of gamma emission rates from Fe and Cu dosimeters extracted from a BWR type reactor vessel were carried out in order to determine their total activity. The dosimeter's activity is related to the neutron flux there by taking into account the reactor material's embrittlement caused by neutron bombardment. The dosimeters were taken out after the first reactor operation cycle. From gamma radioactivity measurements of these dosimeters, neutron flux and fluence were calculated. These parameters are used in the determination of shift and adjusted reference temperature values needed for the development of pressure-temperature curves used during reactor operation

  16. Improvement of stabilizer-based entanglement distillation protocols by encoding operators

    International Nuclear Information System (INIS)

    This paper presents a method for enumerating all encoding operators in the Clifford group for a given stabilizer. Furthermore, we classify encoding operators into the equivalence classes such that EDPs (entanglement distillation protocols) constructed from encoding operators in the same equivalence class have the same performance. By this classification, for a given parameter, the number of candidates for good EDPs is significantly reduced. As a result, we find the best EDP among EDPs constructed from stabilizer codes. This EDP has a better performance than previously known EDPs over a wide range of fidelity

  17. Weak Convergence and Vector-Valued Functions: Improving the Stability Theory of Feynman's Operational Calculi

    International Nuclear Information System (INIS)

    In this paper we present a theorem that establishes a relation between continuous, norm-bounded functions from a metric space into a separable Hilbert space and weak convergence of sequences of probability measures on the metric space. After establishing this result, it's application to the stability theory of Feynman's operational calculi will be illustrated. We will see that the existing time-dependent stability theory of the operational calculi will be significantly improved when the operator-valued functions take their values in LH,H a separable Hilbert space

  18. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    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 speed 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. The accomplishment of detailed and accurate simulations in complex power plants at high speed and low cost are due chiefly to two reasons. The first reason is the application of five distinct modeling principles [2] which are not employed in any other simulation code. The second, and even more important reason is the utilization of a special-purpose peripheral computer with its 13 task-specific parallel processors

  19. 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).

  20. Advanced Construction of Compact Containment BWR

    International Nuclear Information System (INIS)

    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

  1. 44 BWR Waste Package Loading Curve Evaluation

    International Nuclear Information System (INIS)

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

  2. 44 BWR Waste Package Loading Curve Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Scaglione

    2001-11-05

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

  3. New tools for the stability control; Nuevas herramientas para el control de la estabilidad

    Energy Technology Data Exchange (ETDEWEB)

    Melera, J.; Albendea, M.; Escriva, A.; Munoz-Cobos, L. L.; Montesinos

    2006-07-01

    IBERDROLA INGENIERIA y CONSTRUCCION, IBERDROLA GENERACION and the polytechnical Univisty of Valencia, are carrying out a project of R and D called DROP (Decay Ratio on-line Predictor) whose objective is to create a tool (predictor and monitor) for on-line monitoring of the thermal-hydraulic stability during the operation of BWR reactors. DROP is intended to be integrated in CAPRICOPE; the Core Monitoring System of Cofrentes NPP, to allow supervisors to known stability margins during the operation, simplifying reactor start-up and shutdown. In transient conditions, the knowledge of reactor stability will avoid reactor unplanned shutdowns. (Author) 11 refs.

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

    Science.gov (United States)

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

    2014-09-01

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

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

    International Nuclear Information System (INIS)

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

  6. BWR Source Term Generation and Evaluation

    International Nuclear Information System (INIS)

    This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-000061. 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 Quality

  7. Virasoro frames and their Stabilizers for the E_8 lattice type Vertex Operator Algebra

    OpenAIRE

    Griess Jr., Robert L.; Hoehn, Gerald

    2001-01-01

    The concept of a framed vertex operator algebra was studied in [DGH] (q-alg/9707008). This article is an analysis of all Virasoro frame stabilizers of the lattice VOA V for the E_8 root lattice, which is isomorphic to the E_8-level 1 affine Kac-Moody VOA V. We analyze the frame stabilizers, both as abstract groups and as subgroups of the Lie group Aut(V) = E_8(C). Each frame stabilizer is a finite group, contained in the normalizer of a 2B-pure elementary abelian 2-group in Aut(V). In particu...

  8. Comparative analysis of pellet-cladding interaction from IFA-431 and IFA-432 Halden reactor tests. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hann, C.R.; Marshall, R.K.

    1977-07-01

    Two test assemblies containing a total of 12 instrumented fuel rods were irradiated in the HBWR to obtain well-characterized data for fuel operating in the linear heat ranges of commercial nuclear power plants. These data are needed for verification of GAPCON-THERMAL and FRAP computer codes and will provide a series of benchmarks for indexing other thermal performance codes used for reactor safety analysis. Two essentially identical test assemblies, IFA-431 and IFA-432, each containing six instrumented fuel rods, were irradiated under similar conditions. Parameters in the test include pellet-cladding gap size and/or gap eccentricity, fill gas composition, fuel density and stability, linear heat rating, and burnup. The BWR-6 geometry except for length, a 95% theoretical density (TD) pellet, a 0.229 mm (9 mils) diameter gap, and helium fill gas were selected for the reference rod.

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

    International Nuclear Information System (INIS)

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B4C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

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

    International Nuclear Information System (INIS)

    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

  11. Reliability of BWR high pressure core cooling

    International Nuclear Information System (INIS)

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

  12. Control method for BWR type power plant

    International Nuclear Information System (INIS)

    The present invention provides a method of controlling a BWR type plant having internal pumps capable of sufficiently utilizing the performance of a whole volume turbine bypass plant to enable stable supply of electric power upon load interruption of power generator thereof. Namely, upon occurrence of load interruption of a power generator or turbine trip, a plurality of internal pumps are tripped simultaneously to abruptly reduce a reactor core flow rate by a predetermined value or more. In this case, a reactor core flow rate abruptly reduction scram signal is prevented. Alternatively, a plurality of internal pumps are tripped simultaneously to abruptly reduce the reactor core flow rate. In this case, a reactor core flow rate abrupt reduction scram set value is changed in order to inhibit the reactor core flow rate abrupt reduction scram signal. With such procedures, upon load interruption of power generator or upon trip of turbine, reactor core flow rate is abruptly reduced by trip of internal pumps for avoiding increase of neutron fluxes due to reactor pressure change. However, since reactor scram is avoided, the operation can be continued upon load interruption of power generator. As a result, performance of whole volume turbine bypass plant can be utilized sufficiently even upon occurrence of load interruption of power generator. (I.S.)

  13. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A fuel assembly has a 9 x 9 square lattice arrangement having a water channel which occupies an area of 3 x 3 lattice pattern corresponding to 9 fuel rods. Fuel pellets comprise those of not more 7 kinds which have fission products at enrichment degrees different by a spun of not less than 10%. Fuel rods comprise from 4 to 12 first type fuel rods and remaining second type fuel rods. The first type fuel rod is loaded with fuel pellets of fissionable products having an enrichment degree axially different at the upper and the lower portions. The second type fuel rod is loaded with fuel pellets of fissionable products having the same enrichment degree in the vertical direction. With such a constitution, the enrichment degree of fissionable products of fuel pellets in the fuel assembly for a BWR type reactor having different reactor constitution and operation conditions can be used in common. Accordingly, the degree of freedom for the design of the distribution of the enrichment degree is increased. (I.N.)

  14. Hilbert-Huang analysis of BWR neutron detector signals: application to DR calculation and to corrupted signal analysis

    International Nuclear Information System (INIS)

    In this paper, we present an application of the empirical mode decomposition method [Proc. R. Soc. Lond. A 454 (1998) 903], to the stability analysis of BWR. The methodology developed in this paper decomposes the original time series data in intrinsic oscillation modes or IMFs. Then we compute for each IMF, its Hilbert amplitude spectrum and its Hilbert marginal spectrum. From the intrinsic mode related to BWR stability we have obtained by ordinary autoregressive methods the decay ratio value and the oscillation frequency. Also we have proven that the original signal can be reconstructed with seven IMFs and that this modes are mutually orthogonals

  15. Strong and uniform mean stability of cosine and sine operator functions

    Science.gov (United States)

    Sato, Ryotaro; Shaw, Sen-Yen

    2007-06-01

    It is first observed that a uniformly bounded cosine operator function C([dot operator]) and the associated sine function S([dot operator]) are totally non-stable. Then, using a zero-one law for the Abel limit of a closed linear operator, we prove some results concerning strong mean stability and uniform mean stability of C([dot operator]). Among them are: (1) C([dot operator]) is strongly (C,1)-mean stable (or (C,2)-mean stable, or Abel-mean stable) if and only if 0[set membership, variant][rho](A)[union or logical sum][sigma]c(A); (2) C([dot operator]) is uniformly (C,2)-mean stable if and only if S([dot operator]) is uniformly (C,1)-mean stable, if and only if , if and only if , if and only if C([dot operator]) is uniformly Abel-mean stable, if and only if S([dot operator]) is uniformly Abel-mean stable, if and only if 0[set membership, variant][rho](A).

  16. Bivariate empirical mode decomposition applied to the estimation of out-of-phase oscillations in BWR

    International Nuclear Information System (INIS)

    Highlights: • Bivariate empirical mode decomposition (BEMD) in BWR’s instabilities is studied. • The phase determines out-of-phase oscillations in the BWR instability. • The method based in BEMD does not represent a high computational complexity. • The methodology was validated with Nuclear Power Plants stability benchmarks. • The results show that the method contributes to detect out-of-phase oscillations. - Abstract: In this paper a new method based on the bivariate empirical mode decomposition to estimate the phase of regional (out-of-phase) or global (in-phase) modes associated with instabilities in boiling water reactors (BWR), is explored. The proposed method allows decomposing the analyzed signal (constructed from two different Local Power Range Monitors, LPRMs) in different levels or intrinsic mode functions (IMF). The estimation of the phase between these LPRM signals can be achieved by tracking the modes associated to the instability of the BWR and obtaining the cross-correlation function of their corresponding IMF. This phase determines possible out-of-phase oscillations, which play an important role in the BWR instability. The method is relatively simple to implement and it does not represent a high computational complexity. The methodology was tested with simulated signals and validated with two events reported in the Forsmark and Ringhals stability benchmarks. The results of the cases studied show that the proposed method clearly contributes on the fact to detect possible cases of out-of-phase oscillations

  17. Control Rod Pattern Planning of a BWR using Enhanced Nelder-Mead Method

    International Nuclear Information System (INIS)

    We propose a new optimization algorithm for the short-term planning of control rod patterns in an operating BWR. This algorithm is based on the enhanced Nelder-Mead simplex method in which convergence ability is improved for constrained problems in several ways. The main characteristic of this approach is it uses continuous values for the axial positions of control rods. Through calculations in an actual BWR plant, we showed that the new algorithm is effective for automation of short-term planning and reduction of the engineer's workload. (authors)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    The measurements carried out on reactor dosimetry are applied mainly at the study on the effects of the radiation in the materials of the reactor; a 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 tracks manufactured in the ININ is presented, for the environmental monitoring in penetrations around the primary container of the Unit 1 of the Laguna Verde power plant. The monitoring of neutrons carried out with ends of radiological protection, during those operational tests of the reactor. (author)

  1. Prony's method application for BWR instabilities characterization

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, Rogelio, E-mail: rogelio.castillo@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Ramírez, J. Ramón, E-mail: ramon.ramirez@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Alonso, Gustavo, E-mail: gustavo.alonso@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Instituto Politecnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico)

    2015-04-01

    Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred.

  2. Prony's method application for BWR instabilities characterization

    International Nuclear Information System (INIS)

    Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred

  3. Power System Stabilizer Based on Robust H∞ Controller for Low Frequency Operating Range

    Directory of Open Access Journals (Sweden)

    Ali Mohamed Yousef

    2012-05-01

    Full Text Available The aim of study is designed of Power System Stabilizer (PSS based on H4 approach for power system stabilization. The uncertainties in power system modeling and operations are considered at designing of H∞ PSS. The bounds of power system parameters are determined over a wide range of low frequency operating conditions. These bounds are used to design a robust H∞ PSS. A sample power system composed a synchronous generator connected to infinite bus through transmission line is simulated. The digital H∞ PSS can achieve good performance over a wide range of operating conditions. A comparison between power system responses at variety of operating conditions using the proposed H∞ PSS and Linear Quadratic Regulator LQR control have been done. H2 PSS is designed and compared with the proposed controller.

  4. Noise analysis of Forsmark 1 data to investigate BWR core local instability

    International Nuclear Information System (INIS)

    BWR core local instability was experienced at Forsmark 1 (F1) during reactor operation in cycle 16. The event has been studied by applying noise analysis and stability calculations to get insight into the event as well as to identify the cause of local instability. The present report is concerned with noise analysis of data collected during start-up in cycle 17. The results of the current study indicates: The F1 core is quite stable in cycle 17. The max. decay ratio (DR) value of 0.37 was obtained from the stability evaluation of an APRM (average power range monitor) and LPRM (local power range monitor) signals measured at 66% (APRM) of reactor power and 4252 Kg/s (SA-HC) of core flow. Compared with the power profile in cycle 17 (as well as in reactor F2), the core in cycle 16 had an extreme power profile with high power and bottom-shifted axial peak in the core periphery esp. at the four quadrant corners. Such a profile decreases the stability margin in the region. It is a common observation that the DR obtained from APRM tends to be higher than that from LPRM if the global instability mechanism is dominant in the core, and vice versa. The comparison of global and local DR values should be an effective method for detecting local instability during the reactor operation. In order to detect the local instability it is important to evaluate the core stability with sufficient number of LPRMs so as to cover the whole core cross section together with APRMs

  5. Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

    International Nuclear Information System (INIS)

    The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In addition, the effect of ICs on the BWR performance is studied for natural convection flow at lower power and modified riser geometry. The IC is coupled to the steam dome for the steam inlet flow and the Reactor Pressure Vessel (RPV) near the feed water entrance for the condensate return flow. Transient calculations are performed using prescribed pressure set points for the SRVs and given time settings for MSIV closure. The effect of the IC on the forced flow is to reduce the rate of pressure rise and thereby decrease the cycling frequency ofthe SRVS. This is the primary objective of any operating IC in a BWR (e.g. Oyster Creek). The response of the reactor thermal and fission power, steam flow rate, collapsed liquid level, and core average void fraction are found to agree with the trend of pressure. The variations in the case of an active IC can be closely related to the creation of a time lag and changes in the cycling frequency of the SRVS. An analysis for natural convection flow in a BWR indicates that the effect of an IC on its transient performance is similar to that for the forced convection system. In this case, the MSIV closure, has resulted in a lower peak pressure due to the magnitude of reduced power. However, the effect of reduced cycling frequency of the SRV due to the IC, and the time lag between the events, are comparable to that for forced convection

  6. Validation of BWR advanced core and fuel nuclear designs with power reactor measurements

    International Nuclear Information System (INIS)

    Power reactor measurements have been important in validating the reliability, performance characteristics and economics of BWR advanced core and fuel designs. Such measurements go beyond the data obtainable from normal reactor operation and provide detailed benchmark data necessary to verify design and licensing computer design and simulation models. In some cases, such as in the validation of the performance of zirconium barrier pellet-cladding-interaction (PCI) resistant cladding, the BWR power reactor measurements have subjected the advanced fuel design to operating conditions more severe than normal operating conditions, thereby providing nuclear-thermal-mechanical-corrosion performance data for accelerated or extended conditions of operation. In some cases destructive measurements have been carried out on BWR power reactor fuel to provide microscopic and macroscopic data of importance in validating design and licensing analysis methods. There is not uniform agreement among core and fuel designers on the needs for special power reactor core and fuel measurements for validation of advanced designs. The General Electric approach has been to error on the side of extensive, detailed measurements so as to assure reliable performance licensing and economic design and predictive capability. This paper is a summary of some of the validative power reactor measurements that have been carried out on advanced BWR core and fuel designs. Some comparisons of predictions with the data are summarized

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

    International Nuclear Information System (INIS)

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

  8. Operational electrochemical stability of thiophene-thiazole copolymers probed by resonant Raman spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Wade, Jessica; Wood, Sebastian; Kim, Ji-Seon, E-mail: ji-seon.kim@imperial.ac.uk [Department of Physics and Centre for Plastic Electronics, Imperial College London, London SW7 2AZ (United Kingdom); Beatrup, Daniel; Hurhangee, Michael; McCulloch, Iain; Durrant, James R. [Department of Chemistry and Centre for Plastic Electronics, Imperial College London, London SW7 2AY (United Kingdom); Bronstein, Hugo [Department of Chemistry and Centre for Plastic Electronics, Imperial College London, London SW7 2AY (United Kingdom); Department of Chemistry, University College London, London WC1H 0AJ (United Kingdom)

    2015-06-28

    We report on the electrochemical stability of hole polarons in three conjugated polymers probed by resonant Raman spectroscopy. The materials considered are all isostructural to poly(3-hexyl)thiophene, where thiazole units have been included to systematically deepen the energy level of the highest occupied molecular orbital (HOMO). We demonstrate that increasing the thiazole content planarizes the main conjugated backbone of the polymer and improves the electrochemical stability in the ground state. However, these more planar thiazole containing polymers are increasingly susceptible to electrochemical degradation in the polaronic excited state. We identify the degradation mechanism, which targets the C=N bond in the thiazole units and results in disruption of the main polymer backbone conjugation. The introduction of thiazole units to deepen the HOMO energy level and increase the conjugated backbone planarity can be beneficial for the performance of certain optoelectronic devices, but the reduced electrochemical stability of the hole polaron may compromise their operational stability.

  9. Analysis of the operation quality of a stabilized spectrometer for scattered gamma radiation

    International Nuclear Information System (INIS)

    To eliminate errors resulting from parameter instabilities in various units of a scintillator-based γ-spectrometer, a scattered γ-radiation spectrometer was developed. It is stabilized using the peak of total absorption of radiation of operating source. The stabilizer of the energy scale of the spectrometer is a typical automatic control system. The unbalance signal generated by a differential amplifier controls the amplification factor of a photomultiplier by changing the output voltage of its power supply. A dependence of the spectrometer error on the responces of its units is determined analytically, i.e. the transfer ratio of the controlled power supply unit of a photomultiplier, stabilization factor, constant of an integrator, and thresholds of a differential discriminator. The analytic expression describing this dependence has been used while designing a stabilizing system of a measuring probe of a digital radioisotope densitometer of light media

  10. Stability analysis of slot-entry hybrid journal bearings operating with non-newtonian lubricant

    Directory of Open Access Journals (Sweden)

    H.C. Garg

    2015-09-01

    Full Text Available This paper presents theoretical investigations of rheological effects of lubricant on stability parameters of various configurations of slot-entry hybrid journal bearing system. FEM has been used to solve Reynolds equation governing flow of lubricant in bearing clearance space along with restrictor flow equation using suitable iterative technique. The non-Newtonian lubricant has been assumed to follow cubic shear stress law. The stability parameters in terms of stiffness coefficients, damping coefficients, threshold speed and whirl frequency of different configurations of slot-entry hybrid journal bearing have been computed and presented for wide range of external load while operating with Newtonian and Non-Newtonian lubricants. The computed results reveal that variation of viscosity due to non-Newtonian behavior of lubricant affects bearing stability quite significantly. The results are presented in graphical form and logical conclusions are drawn to identify best possible configuration from stability point of view.

  11. Operational electrochemical stability of thiophene-thiazole copolymers probed by resonant Raman spectroscopy

    International Nuclear Information System (INIS)

    We report on the electrochemical stability of hole polarons in three conjugated polymers probed by resonant Raman spectroscopy. The materials considered are all isostructural to poly(3-hexyl)thiophene, where thiazole units have been included to systematically deepen the energy level of the highest occupied molecular orbital (HOMO). We demonstrate that increasing the thiazole content planarizes the main conjugated backbone of the polymer and improves the electrochemical stability in the ground state. However, these more planar thiazole containing polymers are increasingly susceptible to electrochemical degradation in the polaronic excited state. We identify the degradation mechanism, which targets the C=N bond in the thiazole units and results in disruption of the main polymer backbone conjugation. The introduction of thiazole units to deepen the HOMO energy level and increase the conjugated backbone planarity can be beneficial for the performance of certain optoelectronic devices, but the reduced electrochemical stability of the hole polaron may compromise their operational stability

  12. 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)

  13. ARES - a new BWR simulator

    International Nuclear Information System (INIS)

    Coupling the three-dimensional Analytic Function Expansion Nodal (AFEN) nodal model developed within the READY project in 2000-2001 with a four-equation based stationary-state thermalhydraulics module and a new cross section model, a basis has been created for a sophisticated BWR simulator code. Motivation for distilling the efforts into a new simulator, named ARES (AFEN Reactor Simulator), can be summarized in three main points: Stationary-state analyses required by the safety authorities must be independent from the calculations made by the power utilities. Using a different simulator for some calculations is an effective method for obtaining independent results; In order to keep up with the development in the core analysis field, a 'test bench' is required for testing and evaluating new ideas and models. In addition, accuracy of the commercial codes and the models incorporated in them can be evaluated by benchmarking them against the new simulator; and Writing a new program from scratch is potentially a good way to transfer experience from the first generation of Finnish nuclear engineers. It also gives the opportunity to re-evaluate some of the ideas used in the older codes written in times of significantly smaller computer capacity. (orig.)

  14. BWR pipe crack remedies evaluation

    International Nuclear Information System (INIS)

    This paper presents results on: (a) the influence of simulated BWR environments on the stress-corrosion-craking (SCC) susceptibility of Types 304, 316NG, and 347 stainless (SS); (b) fracture-mechanics crack-growth-rate measurements on these materials and weld overlay specimens in different environments; and (c) residual stress measurements and metallographic evaluations of conventional pipe weldments treated by a mechanical-stress-improvement process (MSIP) as well as those produced by a narrow-gap welding procedure. Crack initiation studies on Types 304 and 316NG SS under crevice and non-crevice conditions in 2890C water containing 0.25 ppM dissolved oxygen with low sulfate concentrations indicate that SCC initiates at very low strains (0 in both directions, and then grew at high rate (parallel to the nominal applied load). Residual stress measurements on MSIP-treated weldments and those produced by a narrow-gap welding procedure indicate that these techniques produce compressive stresses over most of the inner surface near the weld and heat-affected zones

  15. Logical model for the control of a BWR turbine

    International Nuclear Information System (INIS)

    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)

  16. Full system decontamination experience in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, N.; Sugai, K.; Katayouse, N.; Fujimori, A.; Iida, K.; Hayashi, K. [Tokyo Electric Power Company, Tokyo (Japan); Kanasaki, T.; Inami, I. [Toshiba Corporation, Yokohama (Japan); Strohmer, F. [Framatome ANP Gmbh, Eelangen (Germany)

    2002-07-01

    At the Fukushima Daiichi Nuclear Power Station unit 3, unit 2, unit 5 and unit 1 of Tokyo Electric Power Company (TEPCO), the replacement of the core shroud and internals has been conducted since 1997 in this order. The welded core internals in operating BWR plants were replaced to improve stress corrosion cracking (SCC) resistance. At present these units are operating smoothly. The developed technology concept is to restore those internals in open air inside the reactor pressure vessel (RPV). To reduce the radiation dose rate inside the RPV, not only a shielding method was applied to cut the radiation from the irradiated structures but also a chemical decontamination method was applied to dissolve the radioactive crud deposited on the surface by using chemical agents. The calculated decontamination factor (DF) at the RPV bottom reached 35-117. As result, the dose rate decreased to approximately 0.1 mSv/h under water. Before and after the installation of the in-vessel shielding, a mechanical cleaning was extensively applied inside the RPV to remove the residual crud as well as the swarf, chips from cutting. As a result, the dose rate at the RPV bottom decreased to ranging from 0.2 to 0.4 mSv/h in air. A working environment for human access, which was better than expected, was established inside the RPV, resulting in 70, 140, 50 and 70 man-Sv (estimated) saving respectively at unit 3 (1F-3), unit 2(1F-2), unit 5(1F-5) and unit 1(1F-1). All four full system decontamination (FSDs) contributed to the successful realization of the core shroud replacement project under the dry condition in RPV.

  17. A qualitative content analysis of global health engagements in Peacekeeping and Stability Operations Institute's stability operations lessons learned and information management system.

    Science.gov (United States)

    Nang, Roberto N; Monahan, Felicia; Diehl, Glendon B; French, Daniel

    2015-04-01

    Many institutions collect reports in databases to make important lessons-learned available to their members. The Uniformed Services University of the Health Sciences collaborated with the Peacekeeping and Stability Operations Institute to conduct a descriptive and qualitative analysis of global health engagements (GHEs) contained in the Stability Operations Lessons Learned and Information Management System (SOLLIMS). This study used a summative qualitative content analysis approach involving six steps: (1) a comprehensive search; (2) two-stage reading and screening process to identify first-hand, health-related records; (3) qualitative and quantitative data analysis using MAXQDA, a software program; (4) a word cloud to illustrate word frequencies and interrelationships; (5) coding of individual themes and validation of the coding scheme; and (6) identification of relationships in the data and overarching lessons-learned. The individual codes with the most number of text segments coded included: planning, personnel, interorganizational coordination, communication/information sharing, and resources/supplies. When compared to the Department of Defense's (DoD's) evolving GHE principles and capabilities, the SOLLIMS coding scheme appeared to align well with the list of GHE capabilities developed by the Department of Defense Global Health Working Group. The results of this study will inform practitioners of global health and encourage additional qualitative analysis of other lessons-learned databases. PMID:25826346

  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. Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1998-01-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  20. Automatic approach to stabilization and control for multi robot teams by multilayer network operator

    Directory of Open Access Journals (Sweden)

    Diveev Askhat

    2016-01-01

    Full Text Available The paper describes a novel methodology for synthesis a high-level control of autonomous multi robot teams. The approach is based on multilayer network operator method that belongs to a symbolic regression class. Synthesis is accomplished in three steps: stabilizing robots about some given position in a state space, finding optimal trajectories of robots’ motion as sets of stabilizing points and then approximating all the points of optimal trajectories by some multi-dimensional function of state variables. The feasibility and effectiveness of the proposed approach is verified on simulations of the task of control synthesis for three mobile robots parking in the constrained space.

  1. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

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

  2. Qualification of the RAMONA-3 code against stability measurements in Ringhals-1

    International Nuclear Information System (INIS)

    The RAMONA-3 BWR system transient time domain code was qualified against the BWR Stability Benchmark defined by the Nuclear Science Committee of the OECD Nuclear Energy Agency (OECD/NEA). Altogether 38 stability test points through 4 cycles of operation in Ringhals-1 were simulated. The RAMONA calculations compared to the measured decay ratios showed a mean deviation of 0.01 and an r.m.s. error of 0.07. For two of the cycles, where the simulations were performed as 'blind tests'. i.e. the measured data were not made available until after reporting the results to the NEA Benchmark Committee, the mean deviation was 0.02 with an r.m.s. error of 0.05. The limit cycle out-of-phase oscillating patterns that appeared in one of the tests were accurately predicted by the code. (author)

  3. BWR pipe crack remedies evaluation

    International Nuclear Information System (INIS)

    This paper presents results on: (a) the influence of simulated BWR environments on the stress-corrosion-cracking (SCC) susceptibility of types 304, 316 NG, and 347 stainless steel (SS), (b) fracture-mechanics crack growth rate measurements on these materials and weld overlay specimens in different environments, and (c) residual stress measurements and metallographic evaluations of conventional pipe weldments treated by a mechanical-stress-improvement process (MSIP) as well as those produced by a narrow-gap welding procedure. Crack initiation studies on types 304 and 316 NG SS under crevice and non-crevice conditions in 2890C water containing 0.25 ppm dissolved oxygen with low sulfate concentrations indicate that SCC initiates at low strains (3%) in the nuclear grade material. Crack growth measurements on fracture-mechanics-type specimens, under low-frequency cyclic loading, show that the type 316 NG steel cracks at a somewhat lower rate (≅ 40%) than sensitized type 304 SS in an impurity environment with 0.25 ppm dissolved oxygen; however, the latter material stops cracking when sulfate is removed from the water. Crack growth in both materials ceases under simulated hydrogen-water chemistry conditions (6 ppb oxygen) even with 100 ppb sulfate present in the water. An unexpected results was obtained in the test on a weld overlay specimen in the impurity environment, viz., the crack grew to the overlay interface at a nominal rate, branched at 900 in both directions, and then grew at a high rate (parallel to the nominal applied load). Residual stress measurements on MSIP-treated weldments and those produced by a narrow-gap welding procedure indicate that these techniques produce compressive stresses over most of the inner surface near the weld and heat-affected zones. (orig.)

  4. 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.

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

    International Nuclear Information System (INIS)

    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

  6. 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.

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

    International Nuclear Information System (INIS)

    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)

  8. Low power level safety management of Finnish BWR

    International Nuclear Information System (INIS)

    Good practices in work coordination and safety management have contributed to short refueling outage duration in Finnish BWR plants. Human and organizational factors are considered especially important in the low paper states, which consist of start-up, shut-down and the outage period itself. This originates from the use of external labour during the outage, the number of both contemporary and sequentially linked human actions and the variety of potential ways the personnel can affect the plant state. While the containment barrier does not exist, more organizational and administrative means have to be used in risk management. To promote the safety further, special studies have been carried out. This paper discusses both the low power mode PSA and the studies of work orientation and competence among the operating staff in Olkiluoto BWR plant. An advanced outage control requires also open-minded consideration of potential risks and the means for their reduction. Good results in low power risk management can be reached only by the involvement of both the plant operating and the maintenance staff. A profound safety management is a prerequisite for safe low power states. (author)

  9. BWR-GALE, Radioactive Gaseous and Liquid Waste Release from BWR

    International Nuclear Information System (INIS)

    1 - Description of program or function: The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment. 2 - Method of solution: GALE calculates expected releases based on 1) standardized coolant activities derived from ANS Standards 18.1 Working Group recommendations, 2) release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, 3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environs, and 4) information received on the operation of nuclear power plants. 3 - Restrictions on the complexity of the problem: The liquid release portion of GALE uses subroutines taken from the ORIGEN (CCC-217) to calculate radionuclide buildup and decay during collection, processing, and storage of liquid radwaste. Memory requirements for this part of the program are determined by the large nuclear data base accessed by these subroutines

  10. Development of core-wide and regional stability test facility, SIRIUS, that simulates void reactivity feedback, and stability evaluation

    International Nuclear Information System (INIS)

    The SIRIUS facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for model point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop, which simulates thermal-hydraulics of a natural circulation BWR. A solid-state, series-regulated power supply, that plays a role of simulation output, was designed to attain fast response speed without loss of accuracy. A noise analysis method was performed to calculate decay ratios from dominant poles of a transfer function by applying AR method to time series of the core inlet flow rate. Experiments were conducted with the SIRIUS facility for the nominal operating condition of 3.13 GWt natural circulation BWR. Channel and regional stability decay ratios were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to evaluate the stability sensitivity of the design parameters such as the power profile on the basis of three-dimensional steady-state analysis, the void reactivity coefficients, the core inlet subcooling, and the thermal conductance of the fuel rod. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

  13. Excitation and Stabilization of Passive Dynamics in Locomotion using Hierarchical Operational Space Control

    OpenAIRE

    Hutter, Marco; Fankhauser, Péter; Gehring, Christian; Bloesch, Michael; Hoepflinger, Mark A.; Siegwart, Roland

    2014-01-01

    This paper describes a hierarchical operational space control (OSC) method based on least square optimization and outlines different ways to reduce the dimensionality of the optimization vector. The framework allows to emulate various behaviors by prioritized task-space motion, joint torque, and contact force optimization. Moreover, a methodology is introduced to partially excite the natural dynamics of the robot by open-loop motor regulation while the entire behavior is stabilized by hierarc...

  14. MATHEMATICAL MODEL OF POWER CONSUMPTION FOR SOME OIL PIPE-LINE SECTIONS WITH POOR OPERATIONAL STABILITY

    Directory of Open Access Journals (Sweden)

    J. N. Kolesnik

    2005-01-01

    Full Text Available Mathematical model of power consumption for technologically completed and non-completed oil pipe-line sections with poor operational stability has been developed on the basis of daily indices concerning oil transportation regimes. The model permits to take into account tendencies in power consumption under various time prediction cycles and ranges of oil freight turnover, changes in the bulk and characteristics of the transported oil, configuration and design parameters of oil pipe-line.

  15. Spray stability of outwards opening pintle injectors for stratified direct injection spark ignition engine operation

    OpenAIRE

    Marchi, A.; Nouri, J. M.; Yan, Y; Arcoumanis, C.

    2010-01-01

    The spray characteristics and spray stability from three prototype piezoelectric pintle-type injectors were investigated under different operating conditions in an optical direct injection engine designed for stratified combustion. The pintle-type outwards opening injector has the potential to address and overcome many of the typical problems related to close-spacing, spray-guided configurations owing to its hollow cone spray, exhibiting better air utilization than multihole sprays, with good...

  16. Computing Stabilized Norms for Quantum Operations via the Theory of Completely Bounded Maps

    OpenAIRE

    Johnston, Nathaniel; Kribs, David W.; Paulsen, Vern I.

    2007-01-01

    The diamond and completely bounded norms for linear maps play an increasingly important role in quantum information science, providing fundamental stabilized distance measures for differences of quantum operations. Based on the theory of completely bounded maps, we formulate an algorithm to compute the norm of an arbitrary linear map. We present an implementation of the algorithm via Maple, discuss its efficiency, and consider the case of differences of unitary maps.

  17. INEL Operable Unit 7-13 containment and stabilization configuration option

    International Nuclear Information System (INIS)

    A containment and stabilization configuration option has been developed for the Idaho National Engineering Laboratory's Subsurface Disposal Area Operable Unit 7-13, the transuranic (TRU)-contaminated waste pits and trenches. The configuration option is presented as an end-to-end system block diagram. Functional subelements are separately discussed, and technical background information, assumptions, input, high-level subelement requirements, and output are presented for each option

  18. Investigation of the Adaptability of Transient Stability Assessment Methods to Real-Time Operation

    OpenAIRE

    Weckesser, Johannes Tilman Gabriel; Jóhannsson, Hjörtur; Sommer, Stefan; Østergaard, Jacob

    2012-01-01

    In this paper, an investigation of the adaptability of available transient stability assessment methods to real-time operation and their real-time performance is carried out. Two approaches based on Lyapunov’s method and the equal area criterion are analyzed. The results allow to determine the runtime of each method with respect to the number of inputs. Furthermore, it allows to identify, which method is preferable in case of changes in the power system such as the integration of distributed ...

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

    International Nuclear Information System (INIS)

    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 m3/hour with cost US$ 0.58/m3. The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  20. 44-BWR WASTE PACKAGE LOADING CURVE EVALUATION

    International Nuclear Information System (INIS)

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

  1. On-line critical control rod pattern prediction algorithm for BWR plant startup

    International Nuclear Information System (INIS)

    This paper describes an on-line algorithm for predicting the critical control rod pattern, which has been developed to reduce the mental strain on operators while withdrawing control rods in the BWR plant startup operation. The proposed algorithm estimates a target eigenvalue (eigenvalue bias) for a three-dimensional neutron kinetics model with a neutron source incorporating actual neutron detector readings. The critical control rod pattern is then predicted based on the estimated eigenvalue bias. The algorithm has been verified using data obtained from an actual startup operation on a BWR model-5 plant, and the estimated eigenvalue bias agreed well with the effective multiplication factor at the criticality actually determined from the operator's judgement. (author)

  2. 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)

  3. BWR lower plenum debris bed models for MELCOR

    International Nuclear Information System (INIS)

    Work is underway at Oak Ridge National Laboratory (ORNL) to incorporate certain models of the Boiling Water Reactor Severe Accident Response (BWRSAR) code into a local version of MELCOR. Specifically, the BWR lower plenum debris bed and bottom head response models taken from BWRSAR are being tested within the local MELCOR code structure. Upon successful completion of testing, recommendations for formal adoption of these models will be made to the Nuclear Regulatory Commission (NRC) and to the MELCOR code development staff at Sandia National Laboratories (SNL). The SNL code development staff retain exclusive responsibility for maintaining the configuration control for the official version of MELCOR. The BWR lower plenum debris bed and bottom head response models permit the calculation of heatup, melting, and relocation of the debris after dryout. They predict the response of the lower plenum internal structures and the bottom head as well as the composition and timing of material release from the vessel. They have been previously applied in severe accident analyses for the Containment Performance Improvement (CPI) Program and the Mark I shell survivability study (NUREG/CR-5423), and in recent assessments of candidate accident management strategies. This paper provides a brief description of the purpose and operation of these models. 11 refs., 15 figs., 5 tabs

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. An ecological interface design for BWR nuclear power plants

    International Nuclear Information System (INIS)

    An ecological interface design was applied to realize the support function for the operator's direct perception and analytical reasoning in the development of an intelligent man-machine system for BWR nuclear power plants. The abstraction-aggregation functional hierarchy representation of the work domain is a base of the ecological interface design. Another base is the concept of the level of cognitive control. The former was mapped into the interface to externalize the operator's normative mental model of the plants, which will reduce his/her cognitive work load and support knowledge-based problem solving. In addition, the same framework can be used for the analytical evaluation of man-machine interfaces. The information content and structure of a prototype interface were evaluated. This approach seems promising from these experiences. (author)

  7. Development of jet pump inspection equipments in BWR

    International Nuclear Information System (INIS)

    This paper describes development of the remotely operated equipments for jet pump ultrasonic testing (UT) in boiling water reactors (BWRs) to enhance the availability of operating nuclear power plants. Stress corrosion cracking (SCC) in the reactor internals has been a major concern in the BWR in recent years. The developed equipments can accomplish the appropriate positioning precision as an application of the Toshiba phased array immersion UT technique and enhance the jet pump inspection performance with a shorter duration and reducing the load for the installation of them. Three types of inspection equipments are developed to cover the outside and inside of the jet pump inlet mixer and the diffuser without disassembling the inlet mixer and the outside of the jet pump riser elbow. Their configurations and specifications are shown in the paper respectively. (author)

  8. BWR plant analyzer development at BNL [Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    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. Operating the plant, quality assurance, and the job of the operating staff, Volume Twelve

    International Nuclear Information System (INIS)

    Subject matter includes operating the plant (the role of the operator, the control room, plant technical specifications, plant operating procedures, initial startup program, BWR/PWR plant startup, BWR/PWR steady state power operation, BWR/PWR transient operation, emergency operation), quality assurance (what is quality, what is quality control, quality assurance includes quality control, government regulation and quality assurance, administrative controls for nuclear power plants, the necessity of reviews and audits, practical quality assurance), and the job of the operating staff (the plant operating staff, plant safety, first aid and resuscitation, general plant hazards, personnel protective equipment, handling chemicals, handling compressed gas, equipment repair and maintenance, communicating with others

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

    International Nuclear Information System (INIS)

    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

  11. Lessons Learned: Employment and Tactical Use of The Combat Medic During Stability Support Operations.

    Science.gov (United States)

    Eldred, Michael S

    2016-01-01

    It is the intent of this article to define the strategy by which Combat Medics have been employed in contemporary stability operations and counterinsurgency conflicts. This article describes the advances in training based on Tactical Combat Casualty Care and how training evolved into an evidence-based model. Training platforms evolved with shifts in mission requirements, new technology, improved medical techniques, and changing protocols. The last portion of this article details recommendations in doctrine, materiel, and training that could enable optimal sustainment standards while retaining operational capability across a wide variety of combat and peace operations. Lessons learned and changes adapted for Medics that are addressed: (1) advances in training and employment of the Combat Medic necessitated by tactics and strategy of current conflicts, (2) Combat Medic regulatory requirements and centralized, just-in-time training, and (3) changes in sustainment training driven by certification requirements and use of medical simulation training centers. PMID:27215893

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

    International Nuclear Information System (INIS)

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

  13. 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

  14. On Stability of Square Root Domains for Non-Self-Adjoint Operators Under Additive Perturbations

    OpenAIRE

    Gesztesy, Fritz; Hofmann, Steve; Nichols, Roger

    2012-01-01

    Assuming $T_0$ to be an m-accretive operator in the complex Hilbert space $\\mathcal{H}$, we use a resolvent method due to Kato to appropriately define the additive perturbation $T = T_0 + W$ and prove stability of square root domains, that is, $$ dom\\big((T_0 + W)^{1/2}\\big) = dom\\big(T_0^{1/2}\\big). $$ Moreover, assuming in addition that $dom\\big(T_0^{1/2}\\big) = dom\\big((T_0^*)^{1/2}\\big)$, we prove stability of square root domains in the form $$dom\\big((T_0 + W)^{1/2}\\big) = dom\\big(T_0^{1...

  15. RF phase stability in the 100-MeV proton linac operation

    Science.gov (United States)

    Seol, Kyung-Tae

    2015-02-01

    The 100-MeV proton linac of the Korea multi-purpose accelerator complex (KOMAC) has been operated to provide a proton beam to users. The 100-MeV linac consists of a 3-MeV radio-frequency quadrupole accelerator (RFQ), four 20-MeV drift-tube linac (DTL) tanks, two medium-energy beam-transmitter (MEBT) tanks, and seven 100-MeV DTL tanks. The requirements of the field stability are within ±1% in RF amplitude and ±1 degree in RF phase. The RF phase stability is influenced by a RF reference line, RF transmission lines, and a RF control system. The RF reference signal is chosen to be a 300-MHz local oscillator (LO) signal, and a rigid copper coaxial line with temperature control was installed for an RF reference distribution. A phase stability of ±0.1 degrees was measured under a temperature change of ±0.1 °C. A digital feedback control system with a field-programmable gate-array (FPGA) module was adopted for a high RF stability. The RF phase was maintained within ±0.1 degrees with a dummy cavity and was within ±0.3 degrees at RFQ operation. In the case of the 20-MeV DTL tanks, one klystron drives 4 tanks, and the input phases of 4 tanks were designed to be in phase. The input phases of 4 tanks were fixed within ±1 degree by adjusting a phase shifter in each waveguide.

  16. Use of scaled BWR lower plenum boron mixing tests to qualify the boron transport model used in TRACG

    International Nuclear Information System (INIS)

    In 2001 GEH applied best estimate methods combined with a statistical methodology to determine upper bound limits for key licensing parameters for anticipated operation occurrence (AOO) transient and anticipated transients without scram (ATWS) overpressure analyses for operating Boiling Water Reactors (BWRs). The methodology was subsequently extended for ESBWR AOO, ATWS, loss of coolant, and stability analyses. GEH is extending the methodology to long-term ATWS analyses for the operating BWRs. A long-term ATWS scenario uses injection of borated water to achieve reactor shutdown. Predicting the mixing and transport of boron is important for calculating the impact on the key licensing parameters. For the many operating BWRs where the denser boron solution is injected into the lower plenum, stratification may occur, delaying boron transport to the core region. CFD modeling can be used to model the stratification and mixing of the boron solution, but such calculations are extremely computer intensive and not cost effective; therefore, a more-empirical approach supported by a theoretical scaling of the dominant phenomena and backed by test data and benchmark calculations is used. The paper presents the TRACG lower plenum boron transport model qualification effort. The scaling basis used to implement the TRACG boron transport model for BWR applications is discussed. (authors)

  17. Stability improvement of an operational two-way satellite time and frequency transfer system

    Science.gov (United States)

    Huang, Yi-Jiun; Fujieda, Miho; Takiguchi, Hiroshi; Tseng, Wen-Hung; Tsao, Hen-Wai

    2016-04-01

    To keep national time accurately coherent with coordinated universal time, many national metrology institutes (NMIs) use two-way satellite time and frequency transfer (TWSTFT) to continuously measure the time difference with other NMIs over an international baseline. Some NMIs have ultra-stable clocks with stability better than 10-16. However, current operational TWSTFT can only provide frequency uncertainty of 10-15 and time uncertainty of 1 ns, which is inadequate. The uncertainty is dominated by the short-term stability and the diurnals, i.e. the measurement variation with a period of one day. The aim of this work is to improve the stability of operational TWSTFT systems without additional transmission, bandwidth or increase in signal power. A software-defined receiver (SDR) comprising a high-resolution correlator and successive interference cancellation associated with open-loop configuration as the TWSTFT receiver reduces the time deviation from 140 ps to 73 ps at averaging time of 1 h, and occasionally suppresses diurnals. To study the source of the diurnals, TWSTFT is performed using a 2  ×  2 earth station (ES) array. Consequently, some ESs sensitive to temperature variation are identified, and the diurnals are significantly reduced by employing insensitive ESs. Hence, the operational TWSTFT using the proposed SDR with insensitive ESs achieves time deviation to 41 ps at 1 h, and 80 ps for averaging times from 1 h to 20 h.

  18. Chapter 3: MHD stability, operational limits and disruptions [Progress in the ITER Physics Basis (PIPB)

    International Nuclear Information System (INIS)

    ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints

  19. Investigation of the Adaptability of Transient Stability Assessment Methods to Real-Time Operation

    DEFF Research Database (Denmark)

    Weckesser, Johannes Tilman Gabriel; Jóhannsson, Hjörtur; Sommer, Stefan;

    2012-01-01

    In this paper, an investigation of the adaptability of available transient stability assessment methods to real-time operation and their real-time performance is carried out. Two approaches based on Lyapunov’s method and the equal area criterion are analyzed. The results allow to determine the...... runtime of each method with respect to the number of inputs. Furthermore, it allows to identify, which method is preferable in case of changes in the power system such as the integration of distributed power resources (DER). A comparison of the performance of the analyzed methods leads to the suggestion...

  20. Development and implementation of a BWR digital feedwater control system

    International Nuclear Information System (INIS)

    EPRI and Northern States Power Company (NSP) realized that fault-tolerant digital technology could improve Feedwater Control System reliability and operations. The Digital Feedwater Control System (DFCS) is the first major fault-tolerant digital control application in a nuclear power plant in the US. The microprocessor-based controller replaced the analog controller in the feedwater control loop to improve performance and reliability of control including ease of maintenance and spare parts supply. The DFCS at Monticello plant has been in operation since July 1986 without any failure of the control system. This system replaced and upgraded the main and start-up analog controllers at Monticello BWR. It features automatic control, on-line signal validation, controller self-diagnosis, and fault tolerance. The dual-redundant hardware configuration minimizes spare parts availability problems. At a control room panel, operators select each feedwater valve's operating mode (one or three element control, manual, and so on) or set bias inputs for individual feed-water-valve demand. These and other features permit more exact feedwater control system tuning, improving feedwater control in all modes of plant operation. Signal validation using parity-space techniques isolated failed sensors and permit system switching to accurate sensors, thus avoiding outages. To ensure successful operation of the system, extensive verification and validation effect were conducted. These included design reviews, factory acceptance testing using simulation code, site acceptance testing using full-scale plant simulator, and pre-operational and operational testing at Monticello plant

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

    International Nuclear Information System (INIS)

    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

  2. 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

  3. Examination of overlay repaired BWR pipe joints

    International Nuclear Information System (INIS)

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

  4. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

    In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

  5. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

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

    International Nuclear Information System (INIS)

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

  7. Stability of LMR oxide pins and blanket rods during run-beyond-cladding-break (RBCB) operation

    International Nuclear Information System (INIS)

    Since 1981, the U.S. Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan have collaborated on an operational reliability testing program in the Experimental Breeder Reactor II. The tests were designed to determine the irradiation behavior of liquid-metal reactor (LMR) oxide pins and blanket rods during steady-state, transient, and run-beyond-claddin-breach (RBCB) operation. Phase I tests completed in 1987 involved current LMR oxide designs and claddings; the phase II tests begun in 1988 concentrate on advanced LMR designs, large-diameter pins (7.5 mm), and advance cladding alloys. The cladding breaches in these tests have been readily detected by fission-gas and delayed-neutron (DN) precursor release. The condition of the fuel pin has been monitored by these releases during RBCB operation. A variety of failures have been intentionally studied in the RBCB portion of the program for operating times of up to 142 full-power days; also, several failure types have been incidentally experienced during the transient tests. Types of failure have included those induced by gas-pressure loading either naturally or by prethinning of the cladding defects, and fuel-cladding mechanical interaction (FCMI)-induced failures or secondary failures caused by the formation of low-density fuel-sodium reaction product (FSRP). This paper summarizes this experience with regard to LMR oxide fuel stability during RBCB operation

  8. 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)

    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 137Cs 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.)

  9. Swedes repair BWR thermal fatigue cracks

    International Nuclear Information System (INIS)

    The discovery of cracks in the feedwater and shutdown cooling systems of Sweden's Barseback 2 BWR in 1980 led to investigations in other Swedish nuclear power stations. Similar cracks were found and the defective parts repaired or replaced before being returned to service. The cause of the cracks has been evaluated and efforts are being made to prevent a recurrence. Experience with Ringhals 1, Orkarsham 2 and Forsmark 1 systems are also described. (author)

  10. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

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

  11. Advanced Neutronics Tools for BWR Design Calculations

    International Nuclear Information System (INIS)

    This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)

  12. Analysis of BWR lattices to recycle americium

    International Nuclear Information System (INIS)

    This study was carried out to assess the ability to eliminate meaningful quantities of americium in a primarily thermal neutron flux by 'spiking' modern BWR fuel with this minor actinide (MA). The studies carried out so far include the simulation of modern 10 x 10 BWR lattices employing the Westinghouse lattice physics code PHOENIX-4 alongside validation studies using MCNP5 models of the same lattices that were spatially depleted via the MONTEBURNS code coupling to ORIGEN. When considering the total inventory of minor actinides in Am-spiked pins, excluding isotopes of uranium and plutonium, the results indicate that a reduction of approximately 50% or more in the total mass inventory of these minor actinides is viable within the selected pins. Therefore, these preliminary results have encouraged the extension of this work to the development of improved lattice designs to help optimize the transmutation rates as well as absolute MA inventory reductions. The ultimate goal being to design batches of these advanced BWR bundles alongside multi-cycle core reload strategies. (authors)

  13. Decontamination techniques for BWR power generation plant

    International Nuclear Information System (INIS)

    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.)

  14. Advanced neutronics tools for BWR design calculations

    International Nuclear Information System (INIS)

    This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy

  15. Rapid dewatering of Powdex resins at a BWR

    International Nuclear Information System (INIS)

    For most BWR's a large portion of their radioactive waste produced is water demineralization resins, both powdered and bead. In order to minimize the quantities of resins produced, proper demineralizer operation, volume reduction and minimization techniques are relevant to spent resin dewatering and packaging. To meet burial requirements spent resin needs to be dewatered and packaged properly. Methods of dewatering spent resins have included centrifuge separation and pulling water out of the resin with a diaphragm pump. Various vendors are offering systems that provide rapid dewatering and volume reduction using filter/liners in combination with vacuum pumps and air blowers. The various systems required a standardized test program for proper comparison and evaluation. The program is described in this paper

  16. Siemens Nuclear Power Corporation experience with BWR and PWR fuels

    International Nuclear Information System (INIS)

    The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

  17. Influence of local regulations on TN dual purpose BWR casks

    International Nuclear Information System (INIS)

    Transnucleaire (Paris, France) and Transnuclear, Inc. (Hawthorne, New York, United Sates) have both developed Pressurized Water Reactor (PWR) spent fuel casks for storage and transport purposes. The products are supplied in Europe by Transnucleaire and in the United States by Transnuclear, Inc. Now the TN Group is working on a design for Boiling Water Reactor (BWR) spent fuel assemblies: the TN 52 L cask is designed for transport and storage is Switzerland, the TN 68 cask is designed for transport and storage in the United States. For storage purpose, national regulatory requirements have to be met: each country has specific demands and criteria. As a consequence, differences between the TN 52 L design and the TN 68 design for rather similar contents appear in several fields: the design work, the licensing process, the manufacturing and the operational life. (author)

  18. Dynamic safety systems in BWR plant safety systems

    International Nuclear Information System (INIS)

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

  19. Control method for water quality of BWR type reactor

    International Nuclear Information System (INIS)

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

  20. Analysis of BWR OPRM plant data and detection algorithms with DSSPP

    International Nuclear Information System (INIS)

    All U.S. BWRs are required to have licensed stability solutions that satisfy General Design Criteria (GDC) 10 and 12 of 10 CFR 50 Appendix A. Implemented solutions are either detect and suppress or preventive in nature. Detection and suppression of power oscillations is accomplished by specialized hardware and software such as the Oscillation Power Range Monitor (OPRM) utilized in Option III and Detect and Suppress Solution - Confirmation Density (DSS-CD) stability Long-Term Solutions (LTSs). The detection algorithms are designed to recognize a Thermal-Hydraulic Instability (THI) event and initiate control rod insertion before the power oscillations increase much higher above the noise level that may threaten the fuel integrity. Option III is the most widely used long-term stability solution in the US and has more than 200 reactor years of operational history. DSS-CD represents an evolutionary step from the stability LTS Option III and its licensed domain envelopes the Maximum Extended Load Line Limit Analysis Plus (MELLLA +) domain. In order to enhance the capability to investigate the sensitivity of key parameters of stability detection algorithms, GEH has developed a new engineering analysis code, namely DSSPP (Detect and Suppress Solution Post Processor), which is introduced in this paper. The DSSPP analysis tool represents a major advancement in the method for diagnosing the design of stability detection algorithms that enables designers to perform parametric studies of the key parameters relevant for THI events and to fine tune these system parameters such that a potential spurious scram might be avoided. Demonstrations of DSSPPs application are also presented in this paper utilizing actual plant THI data. A BWR/6 plant had a plant transient that included unplanned recirculation pump transfer from fast to slow speed resulting in about 100% to ∼40% rated power decrease and about 99% to ∼30% rated core flow decrease. As the feedwater temperature is reduced

  1. Oxidation of lactose to lactobionic acid by a Microdochium nivale carbohydrate oxidase: kinetics and operational stability.

    Science.gov (United States)

    Nordkvist, Mikkel; Nielsen, Per Munk; Villadsen, John

    2007-07-01

    Oxidation of lactose to lactobionic acid by a Microdochium nivale carbohydrate oxidase was studied. The K(m)-value for lactose, obtained by a traditional enzymatic assay, was 0.066 mM at pH 6.4 and 38 degrees C. The effect of oxygen on the enzymatic rate of reaction as well as the operational stability of the enzyme was studied by performing reactions at constant pH and temperature in a stirred tank reactor. Catalase was included in all reactions to avoid inhibition and deactivation of the oxidase by hydrogen peroxide. At pH 6.4 and 38 degrees C, K(m) for oxygen was 0.97 mM, while the catalytical rate constant, k(cat), was 94 s(-1). Furthermore, we found that the operational stability of the oxidase was dependent on the type of base used for neutralization of the acid produced. Thus, when 2 M NaOH was used for neutralization of a reaction medium containing 50 mM phosphate buffer, significant deactivation of the oxidase was observed. Also, we found that the oxidase was protected against deactivation by base at high lactose concentrations. A simple model is proposed to explain the obtained results. PMID:17154316

  2. Development of the neutron source evaluation method and predictor of SRM/SRNM count rate in BWR simulator

    International Nuclear Information System (INIS)

    The source range monitors (SRMs) and the start-up range neutron monitors (SRNMs) are important instruments from the BWR criticality safety viewpoints. There is a limitation of the minimum count rate (3cps) to guarantee the normality of the SRMs/SRNMs. After the long outage, this limitation is critical for the fuel shuffling due to the decay of the neutron sources in the fuel. The neutron source intensity evaluation method based on a micro burn-up model and the predictor function of the SRM/SRNM count rate are developed in AETNA01, GNF's three-dimensional neutronic-thermal hydraulic boiling water reactor (BWR) core simulator. These new functions are validated through the comparisons between operating BWR's measured data after shutdown and during shuffling. Through these comparisons, high accuracy of the SRM/SRNM count rate predictor of AETNA01 was presented. (author)

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

    International Nuclear Information System (INIS)

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

  4. Operational stability enhancement in organic light-emitting diodes with ultrathin Liq interlayers

    Science.gov (United States)

    Tsang, Daniel Ping-Kuen; Adachi, Chihaya

    2016-03-01

    Organic light-emitting diodes (OLEDs) under constant current operation suffer from a decrease of luminance accompanied by an increase of driving voltage. We report a way to greatly improve the stability of OLEDs having a green emitter exhibiting thermally activated delayed fluorescence (TADF), (4s,6s)-2,4,5,6-tetra(9H-carbazol-9-yl) isophthalonitrile (4CzIPN), by introducing ultrathin (1 to 3 nm) interlayers of 8-hydroxyquinolinato lithium (Liq) between hole-blocking layer and its surrounding emissive and electron-transport layers. Under constant current operation starting at a luminescence of 1,000 cd/m2, the time to reach 90% of initial luminance (LT90) increased eight times, resulting in LT90 = 1,380 hours after insertion of the interlayers. Combining this new concept and mixed host system, LT95 was further extended to 1315 hours that is 16 times of reference device. This is the best value reported for TADF-based OLEDs and is comparable to the operational lifetimes of well-established phosphorescence-based OLEDs. Thermally stimulated current measurements showed that the number of deep charge traps was reduced with the insertion of the ultrathin Liq interlayer, indicating that reducing the number of deep traps is important for improving the operational lifetime and that exciton-polaron annihilation may be a source of the device degradation.

  5. Range of the radiation monitor for the rigid vent of primary containment during normal and emergency operation for a BWR-5 in Laguna Verde; Rango del monitor de radiacion para el venteo rigido de la contencion primaria durante operacion normal y emergencia para un reactor BWR-5 en Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Pozos S, A. M.; Cabrera U, S.; Mata A, J. A.; Sandoval V, S.; Ovando C, R.; Vargas A, A.; Gallardo R, I.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    The earthquake followed by a tsunami, happened in March, 2011 in the coasts of oriental Japan, caused damages in the nuclear power plants 1 at 4 of Fukushima Daiichi leading to damage of the fuel in three of the reactors and to the radiation liberation to the exterior. As consequence of those events, the regulations requires that the power plants with Primary Containment type Mark I and II evaluate to have a system of rigid vent with a monitoring equipment of radiation effluents. The present work covers the rigid vent of diameter 12 of the Primary Containment, type Mark-II, of nuclear power plant of Laguna Verde in conditions of severe accident and normal operation, low regime of Extended Power Up rate (EPU - 2317 MWt), using the codes MAAP3B, MICROSHILED 5.05 and the Bardach Black Boxes methodology. As a result the measurement range of the radiation monitor that is required for monitoring the gassy liberation to the atmosphere was determined. The conclusion is that the superior limit of the range of the radiation meter during a Severe Accident is of 8.55 E + 05 R/h (8.55 E + 08 m R/h) and the superior limit in normal operation of 1.412 E-11 at 2.540 E-7 R/h (1.412 E-14 at 2.540 E-10 m R/h). (Author)

  6. A Deterministic/probalistic analysis of Ex-Vessel melt risk in a BWR

    OpenAIRE

    Abal López, Javier

    2006-01-01

    The present study is concerned with deterministic and probabilistic analysis of ex-vessel melt risks in a Swedish designed BWR plant. The focus is placed on a station blackout (SBO) scenario, with immediate SCRAM and subsequent activation of the main steam valve isolation (at 52 s). Four sequences were examined in detail to study the effect of two valves systems related to the operation of ADS (Automatic Depressurization System), and cavity flooding by water from suppression po...

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

    International Nuclear Information System (INIS)

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

  8. MOX fuel use in a BWR with extended power up-rate

    International Nuclear Information System (INIS)

    Highlights: ► Use of MOX fuel is assessed for a BWR under a extended power uprate (EPU). ► EPU conditions reduce the maximum amount of MOX fuel to be loaded. ► The use of MOX fuel affects mainly the core neutronics and not to the thermal hydraulics. ► Start up of an equilibrium mixed UO2–MOX core under EPU does not present stability problems. -- Abstract: Although MOX fuel coming from reprocessed depleted uranium fuels has been used as a recycling strategy by countries like France and Japan it is not a common policy in the 30 countries that uses nuclear power, nowadays it seems to be a more direct alternative to reduce the depleted fuel interim storage. Previously, the spent fuel pools of Laguna Verde Nuclear Power plant were redesigned to host the total operating life depleted fuel under its original nominal power condition, however the plant has been up-rated to 120% of its original nominal power increasing the number of depleted fuel forecasted. This new situation makes necessary the analysis of alternatives, being one of them recycling. The current paper assesses the viability of using MOX fuel in the up-rated Power Plant; the design of the boiling water reactor MOX fuel addresses the two main constraints of its use: shutdown margin and reactor stability. Fuel design proposed sets the appropriate MOX enrichment and the maximum MOX fuel batch reload that does not imply any modification to the reactor control systems to avoid an extra economical cost due to its use.

  9. Genusa Bepu methodologies for the safety analysis of BWRs; Metodologias Bepu de Genusa para el analisis de seguridad de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trueba, M.; Garcia, J.; Goodson, C.; Ibarra, L.

    2016-08-01

    This article describes the BEPU methodologies developed by General Electric-Hitachi (GEH) for the evaluation of the BWR reactor safety analysis based on the TRACG best-estimate code. These methodologies are applicable to a wide range of events, operational transients (AOO), anticipated transients without scram (ATWS), loss of coolant accidents (LOCA) and instability events; to different BWR types operating commercially. General Electric (GE( designs and other vendors, including Generation III+ESBWR; to the new operation strategies, and to all types of BWR fuel. Their application achieves, among other benefits, a better understanding of the overall plant response and an improvement in margins to the operating limits; thus, the increase of flexibility in reactor operation and reduction in generation costs. (Author)

  10. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

    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

  11. Assessment of severe accident prevention and mitigation features: BWR, Mark III containment design

    International Nuclear Information System (INIS)

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark III containments (BWR Mark II's), have been identified. These features and actions were developed specifically for the Grand Gulf Plant and from assessments of 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 III to severe accident containment loads were also identified. In addition, those features of a BWR Mark III, 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. The 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 Grand Gulf and other Mark III 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

  12. TARMS2, New-generation BWR core management system

    International Nuclear Information System (INIS)

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

  13. Operational stabilities of different chemical derivatives of Novozym 435 in an alcoholysis reaction.

    Science.gov (United States)

    Villalba, María; Verdasco-Martín, Carlos M; Dos Santos, Jose C S; Fernandez-Lafuente, Roberto; Otero, Cristina

    2016-08-01

    Industrial use of Novozym 435 in synthesis of structured lipids and biodiesel via alcoholysis is limited by mass transfer effects of the glycerides through immobilized enzymes and its low operational stability under operation conditions. To better understand this, differently modified Novozym 435 preparations, differing in their surface nature and in their interactions with reactants, have been compared in the alcoholysis of Camelina sativa oil. The three modifications performed have been carried out under conditions where all exposed groups of the enzyme have been modified. These modifications were: 2,4,6-trinitrobenzensulfonic acid (Novo-TNBS), ethylendiamine (Novo-EDA) and polyethylenimine (Novo-PEI). Changes in their operational performance are analyzed in terms of changes detected by scan electron microscopy in the support morphology. The hydrophobic nature of the TNBS accelerates the reaction rate; t-ButOH co-solvent swells the macroporous acrylic particles of Lewatit VP OC 1600 in all biocatalysts, except in the case of Novo-PEI. This co-solvent only increases the maximal conversions obtained at 24h using the modified biocatalysts. t-ButOH reduces enzyme inactivation by alcohol and water. In a co-solvent system, these four biocatalysts remain fully active after 14 consecutive reaction cycles of 24h, but only Novo-TNBS yields maximal conversion before cycle 5. Some deposits on biocatalyst particles could be appreciated during reuses, and TNBS derivatization diminishes the accumulation of product deposits on the catalyst surface. Most particles of commercial Novozym(®) 435 are broken after operation for 14 reaction cycles. The broken particles are fully active, but they cause problems of blockage in filtration operations and column reactors. The three derivatizations studied make the matrix particles more resistant to rupture. PMID:27241290

  14. Mine stability evaluation of panel 1 during waste emplacement operations at WIPP

    International Nuclear Information System (INIS)

    The specific objectives of the work were defined by the Environmental Evaluation Group (EEG) as follows: (1) assess the stability of panel 1 during the proposed operation of waste emplacement; (2) estimate the amount of time before room closure would be expected to transfer rock loads to the waste packages. The work consisted of (1) an analysis of geotechnical data and a review of the Department of Energy's (DOE) plans for waste emplacement in panel 1, (2) an evaluation of ground conditions based on data analysis and observations of changes in ground conditions since the first evaluation in 1993 (USBM 1993), and (3) preparation of a report and presentation of the results to EEG staff. Excluded from this study are radiological safety issues and policies. The study is based on data provided by DOE and Westinghouse Electric Corporation (operator of the site) and conversations with DOE and Westinghouse personnel. MTI cannot independently verify the accuracy of the data within the scope of this study and recommends independent evaluations of data gathering, quality assurance procedures, and structural designs. The operator has the ultimate responsibility for structural designs and has expressed a strong commitment to ensuring worker safety

  15. Operational Support for Instrument Stability through ODI-PPA Metadata Visualization and Analysis

    Science.gov (United States)

    Young, M. D.; Hayashi, S.; Gopu, A.; Kotulla, R.; Harbeck, D.; Liu, W.

    2015-09-01

    Over long time scales, quality assurance metrics taken from calibration and calibrated data products can aid observatory operations in quantifying the performance and stability of the instrument, and identify potential areas of concern or guide troubleshooting and engineering efforts. Such methods traditionally require manual SQL entries, assuming the requisite metadata has even been ingested into a database. With the ODI-PPA system, QA metadata has been harvested and indexed for all data products produced over the life of the instrument. In this paper we will describe how, utilizing the industry standard Highcharts Javascript charting package with a customized AngularJS-driven user interface, we have made the process of visualizing the long-term behavior of these QA metadata simple and easily replicated. Operators can easily craft a custom query using the powerful and flexible ODI-PPA search interface and visualize the associated metadata in a variety of ways. These customized visualizations can be bookmarked, shared, or embedded externally, and will be dynamically updated as new data products enter the system, enabling operators to monitor the long-term health of their instrument with ease.

  16. Development of methodology for early detection of BWR instabilities

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  18. Investigation of burnup credit implementation for BWR fuel

    International Nuclear Information System (INIS)

    Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumptions made to define the axial profile of the burnup and void fraction (for BWR), to determine the composition of the irradiated fuel and to compute the criticality simulation. In the framework of Burnup Credit implementation for BWR fuel, this paper proposes to investigate part of these items. The studies presented in this paper concern: the influence of the burnup and of the void fraction on BWR spent fuel content and on the effective multiplication factor of an infinite array of BWR assemblies. A code-to-code comparison for BWR fuel depletion calculations relevant to Burnup Credit is also performed. (authors)

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

    International Nuclear Information System (INIS)

    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.

  20. 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.

  1. BWR core shroud replacement. A cost-effective alternative

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking of stainless steel internals of BWRs is a longstanding and recognized problem. Well-engineered approaches have been developed by utilities and engineering contractors to address many of the known problems. One of these is cracking of the horizontal and vertical welds in the core shrouds. The approach being used world wide to address cracking in the horizontal welds is to install vertical tie rods and supplemental lateral restraints which provide redundant vertical and lateral support of the shroud and thereby structurally replace the function of the shroud horizontal welds. This approach has been thoroughly analyzed, approved by regulatory bodies and implemented in the United States, Europe, and Taiwan in over 16 BWR units. Importantly, many of these tie rod installations have been installed on a preemptive basis as a prudent preventive maintenance action. As BWR plants age, additional cracking of core shrouds, core supports, core spray piping and jet pumps has been experienced at some plants, indicating the need for a cost-effective, technically-acceptable approach for dealing with these problems. One such approach is complete replacement of the core shroud assembly, including core supports and attached core spray piping. This operation also makes the jet pumps accessible for replacement if this is determined to be necessary. A complete replacement by welding has been accomplished in Japan, and provides one acceptable replacement method. An alternative replacement concept, described in this paper, takes advantage of the proven tie rod modification to provide an approach which can be accomplished without in-reactor welding and without draining the reactor vessel. As described in this paper, this approach can be accomplished with significantly reduced radiation exposure to workers, reduced costs, and in a fraction of the time required for the fully-welded approach. An additional benefit is that as-welded joints in the reactor

  2. A very low offset voltage auto-zero stabilized CMOS operational amplifier

    International Nuclear Information System (INIS)

    A high precision operational amplifier has been developed in a standard .8μ CMOS process. A continuous time auto- zero stabilized architecture was used, that leads to a typical input offset voltage less than 2μV -100 nV/deg. C. The amplifier with its output buffer consumes 5 mW at a supply voltage of ± 2.5 V. The gain bandwidth product is 2 MHz while the slew rate is - 6 V/μS and + 8.8 V/μS, on 10 pF with 10 KΩ load, respectively. This amplifier is suitable for the control of large dynamic (>105) calibration signal, and for very low signal instrumentation. (authors)

  3. The Japanese utilities' requirements for a next century BWR

    International Nuclear Information System (INIS)

    This paper reports on the progress of studies to establish a plant concept for a Boiling Water Reactor (BWR) of the next century. The studies were initiated in 1990 by the Japanese utilities, jointly with NSSS vendors, to investigate evolutionary and long term nuclear power plants. The plant concept is based on the evolution of the ABWR taking advantage of new technology. Fundamental plant philosophies are expressed by the following four desired characteristics: Economical, Benign to human, Simple, Flexible. According to these philosophies, concrete objectives of the plant design are reduction of operating burden and maintenance, increase of safety margin and flexibility to adjust to possible changes in economic circumstances in the years to come. The basic utilities' requirements for the new generation BWR were discussed based on the future social needs and the current operational experiences. Start of operation is to be in the 2010's when the early generation LWRs may need to be replaced. Plant power generation capacity will be about 1500 MWe since this level rating will be achievable by extrapolation of current technology. One important requirement is to achieve power generation costs competitive with other generation methods. An outline of the utilities' requirements follows: Operability; prevent inadvertent reactor scram and engineering safety system actuation due to single failure of normal duty systems or single operator error, achieve same load following capability as ABWR, design for plant availability of up to 90%, achieve plant design life of 60 years, maintain annual inspection period at less than 40 days, reduce maintenance activities in harsh environments, reduce employees' dose to less than that of ABWR, consider 'N+2' design to reduce peak loads during annual inspection. Safety margin; increase grace period for transient and accident events, adopt severe accident countermeasures, keep core damage frequency lower than that of ABWR and conditional

  4. Beam extraction and high stability operation of high current electron cyclotron resonance proton ion source

    Energy Technology Data Exchange (ETDEWEB)

    Roychowdhury, P., E-mail: pradipr@barc.gov.in; Mishra, L.; Kewlani, H.; Mittal, K. C. [Accelerator and Pulse Power Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Patil, D. S. [Laser and Plasma Technology Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2014-03-15

    A high current electron cyclotron resonance proton ion source is designed and developed for the low energy high intensity proton accelerator at Bhabha Atomic Research Centre. The plasma discharge in the ion source is stabilized by minimizing the reflected microwave power using four stub auto tuner and magnetic field. The optimization of extraction geometry is performed using PBGUNS code by varying the aperture, shape, accelerating gap, and the potential on the electrodes. While operating the source, it was found that the two layered microwave window (6 mm quartz plate and 2 mm boron nitride plate) was damaged (a fine hole was drilled) by the back-streaming electrons after continuous operation of the source for 3 h at beam current of 20–40 mA. The microwave window was then shifted from the line of sight of the back-streaming electrons and located after the water-cooled H-plane bend. In this configuration the stable operation of the high current ion source for several hours is achieved. The ion beam is extracted from the source by biasing plasma electrode, puller electrode, and ground electrode to +10 to +50 kV, −2 to −4 kV, and 0 kV, respectively. The total ion beam current of 30–40 mA is recorded on Faraday cup at 40 keV of beam energy at 600–1000 W of microwave power, 800–1000 G axial magnetic field and (1.2–3.9) × 10{sup −3} mbar of neutral hydrogen gas pressure in the plasma chamber. The dependence of beam current on extraction voltage, microwave power, and gas pressure is investigated in the range of operation of the ion source.

  5. Beam extraction and high stability operation of high current electron cyclotron resonance proton ion source

    International Nuclear Information System (INIS)

    A high current electron cyclotron resonance proton ion source is designed and developed for the low energy high intensity proton accelerator at Bhabha Atomic Research Centre. The plasma discharge in the ion source is stabilized by minimizing the reflected microwave power using four stub auto tuner and magnetic field. The optimization of extraction geometry is performed using PBGUNS code by varying the aperture, shape, accelerating gap, and the potential on the electrodes. While operating the source, it was found that the two layered microwave window (6 mm quartz plate and 2 mm boron nitride plate) was damaged (a fine hole was drilled) by the back-streaming electrons after continuous operation of the source for 3 h at beam current of 20–40 mA. The microwave window was then shifted from the line of sight of the back-streaming electrons and located after the water-cooled H-plane bend. In this configuration the stable operation of the high current ion source for several hours is achieved. The ion beam is extracted from the source by biasing plasma electrode, puller electrode, and ground electrode to +10 to +50 kV, −2 to −4 kV, and 0 kV, respectively. The total ion beam current of 30–40 mA is recorded on Faraday cup at 40 keV of beam energy at 600–1000 W of microwave power, 800–1000 G axial magnetic field and (1.2–3.9) × 10−3 mbar of neutral hydrogen gas pressure in the plasma chamber. The dependence of beam current on extraction voltage, microwave power, and gas pressure is investigated in the range of operation of the ion source

  6. BWR 90: An evolutionary ABWR plant for the next decade(s)

    International Nuclear Information System (INIS)

    The future of nuclear power is determined by ''conflicting'' considerations. It must be generated at acceptable cost with negligible environmental effects, and it must achieve and maintain a reasonable level of public acceptance. The latter is, however, not necessarily governed by rational assessments. The ABB approach to ''nuclear'' can be characterized as a ''cautious evolution''. In the next decade(s), its offerings will largely be based on ''evolutions'' of two successful LWR types, the BWR 90 and System 80+. Both designs can be designed, licensed and constructed in accordance with any safety regulations in force or envisaged in the Western world. This paper elaborates on BWR 90, a design that is closely based on its forerunner, taking into account experiences from design and engineering, construction and commissioning, and operation, needs for adapting to new technologies and new safety requirements, as well as possibilities for simplifications and cost savings

  7. Local instability in BWR reactor simulator

    International Nuclear Information System (INIS)

    ''Local'' oscillations in thermal-hydraulic quantities have been observed while performing system code calculations related to PIPER-ONE BWR simulator; void fraction oscillations were detected at the channel middle elevation but not at channel bottom and top. This paper deals with an experimental investigation of ''local'' instability starting from experimental data obtained with the PIPER-ONE facility. The considered test is PO-SD-5B; the post-test analysis has been carried out by RELAP5/MOD2 code. In particular the analysis concerns the experimental and calculated oscillations of pressure drop in the core region and the calculated oscillations of channel void fraction

  8. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

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

  9. Fracture assessment of a BWR pump nozzle

    International Nuclear Information System (INIS)

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

  10. Thermal hydraulic test apparatus to develop advanced BWR fuel bundles with spectral shift rods (SSR)

    International Nuclear Information System (INIS)

    An advanced water rod (WR) called the spectral shift rod (SSR), which replaces a conventional WR in a BWR fuel bundle, enhances the BWR's merit of uranium saving through the spectral shift operation. The SSR consists of an inlet hole, a wide ascending path, a narrow descending path and an outlet hole. The inlet hole locates below a lower tie plate (LTP) and the outlet hole is set above it. In the SSR, water boils by neutron and gamma-ray heating and water level is formed in the ascending path. This SSR water level can be controlled by core flow rate, which amplifies core void fraction change, resulting in the amplified spectral shift effect. Steady state and transient tests were conducted to evaluate SSR thermal-hydraulic characteristics under BWR operation condition. The several types of SSR configuration were tested, which covers SSR design in both next generation and conventional BWRs. In this paper, the test apparatus overview and measurement systems especially two phase water level measures in the SSR are presented. (author)

  11. Painting within BWR secondary containment

    International Nuclear Information System (INIS)

    Paint solvents deposited on charcoal adsorbers used in the standby gas treatment (SBGT) system can reduce the methyl iodide removal efficiency of the adsorbers. The U.S. Nuclear Regulatory Commission (NRC) has imposed regulations in the form of technical specifications in order to mitigate the consequences of potential charcoal adsorber poisoning. Specifically, Monticello Nuclear Generating Plant's technical specifications require that the filters be tested following painting, fire or chemical release in any ventilation zone communicating with the system while the system is operating that could contaminate the HEPA filters or charcoal adsorbers. A method has been developed to determine the maximum allowable painting rate within secondary containment such that the quantity of organic solvents in the building at any time is always less than that which could potentially reduce the charcoal adsorber removal efficiency below the technical specification limit of 90%. This method allows many types of paints to be used within containment, although maximum allowable painting rates will decrease for paints with high quantities of hydrocarbon solvents or slow drying times. Controls in the form of a painting procedure have been put in place to ensure the analysis remains valid during all painting projects. Strict control over painting within secondary containment will ensure the integrity of the charcoal adsorbers and thus enhance the safe operation of the plant

  12. USING THE SULFUR POLYMER STABILIZATION SOLIDIFICATION PROCESS TO TREAT RESIDUAL MERCURY WASTES FROM GOLD MINING OPERATIONS.

    Energy Technology Data Exchange (ETDEWEB)

    BOWERMAN,B.ADAMS,J.KALB,P.WAN,R.Y.LEVIER,M.

    2003-02-24

    Large quantities of mercury are generated as a by-product during the processing of gold ore following mining operations. Newmont Mining Corporation (NMC), which operates some of the world's largest gold mines, sought a method to permanently ''retire'' its mercury by-products, thereby avoiding potential environmental liability. Sulfur Polymer Stabilization-Solidification (SPSS) is an innovative technology developed at Brookhaven National Laboratory (BNL) for treatment of mercury and mercury contaminated materials, such as soil, sludge and debris. BNL conducted a treatability study to determine the potential applicability of SPSS for treatment of Newmont mercury, and the treated product passed the U.S. Environmental Protection Agency (EPA) test for toxicity. The SPSS process has been shown to be effective on radioactive and nonradioactive mercury and mercury-contaminated materials with a pilot-scale batch system capable of producing 0.03 m{sup 3} (1 ft{sup 3}) per batch. Engineering scale-up issues are discussed and material property tests addressing these issues are described.

  13. On Stability of Open-Loop Operation without Rotor Information for Brushless DC Motors

    Directory of Open Access Journals (Sweden)

    Zhong Wu

    2014-01-01

    Full Text Available Open-loop operation mode is often used to control the Brushless DC Motors (BLDCMs without rotor position sensors when the back electromotive force (EMF is too weak due to the very low rotor velocity. The rotor position information is not necessary in this mode and the stator windings are supplied with voltages under a certain ratio of the amplitude to the frequency. However, the rotor synchronization will be destroyed once if the commutation instant is inappropriate. In order to improve the reliability of the open-loop operation mode, a dynamic equation is established to represent the synchronization error between the rotor and the stator. Thereafter, the stability of the open-loop control mode is analyzed by using Lyapunov indirect method. Theoretical analysis indicates that the open-loop control mode is asymptotically stable only when the commutation instant of the stator current lags behind the ideal one suitably. Finally, theoretical analysis is verified through the experimental results of a certain BLDCM.

  14. Impact of MSWI bottom ash codisposed with MSW on landfill stabilization with different operational modes.

    Science.gov (United States)

    Li, Wen-Bing; Yao, Jun; Malik, Zaffar; Zhou, Gen-Di; Dong, Ming; Shen, Dong-Sheng

    2014-01-01

    The aim of the study was to investigate the impact of municipal solid waste incinerator (MSWI) bottom ash (BA) codisposed with municipal solid waste (MSW) on landfill stabilization according to the leachate quality in terms of organic matter and nitrogen contents. Six simulated landfills, that is, three conventional and three recirculated, were employed with different ratios of MSWI BA to MSW. The results depicted that, after 275-day operation, the ratio of MSWI BA to fresh refuse of 1 : 10 (V : V) in the landfill was still not enough to provide sufficient acid-neutralizing capacity for a high organic matter composition of MSW over 45.5% (w/w), while the ratio of MSWI BA to fresh refuse of 1 : 5 (V : V) could act on it. Among the six experimental landfills, leachate quality only was improved in the landfill operated with the BA addition (the ratio of MSWI BA to fresh refuse of 1 : 5 (V : V)) and leachate recirculation. PMID:24779006

  15. Ger-type and Hyers-Ulam stabilities for the first-order linear differential operators of entire functions

    Directory of Open Access Journals (Sweden)

    Sin-Ei Takahasi

    2004-05-01

    Full Text Available Let h be an entire function and Th a differential operator defined by Thf=f′+hf. We show that Th has the Hyers-Ulam stability if and only if h is a nonzero constant. We also consider Ger-type stability problem for |1−f′/hf|≤ϵ.

  16. DESIGN OF ROBUST TWO-AXIS SYSTEMS FOR STABILIZATION AND TRACKING OF INFORMATION-MEASURING DEVICES OPERATED ON GROUND VEHICLES

    Directory of Open Access Journals (Sweden)

    Olga Sushchenko

    2014-06-01

    Full Text Available Design features of the information-measuring robust stabilization and tracking systems operated on the ground vehicles are considered. The mathematical description of the control object mounted in the two-axis gimbals    is obtained. Design features of the two-axis robust stabilization and tracking systems based on the structural H.. -synthesis are researched. The simulation results are given

  17. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

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

  18. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

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

  19. Vibration characteristics of BWR primary containment vessel

    International Nuclear Information System (INIS)

    This paper reports on a part of a seismic proving test of a BWR Primary Containment Vessel (PCV), a vibration analysis that is carried out considering the effect of reinforcement around attached masses such as equipment hatches and a personnel airlock. A deflected shape with the local attached masses and reinforcement is expanded using free vibration modes of an axisymmetric PCV without attached masses and reinforcement. The free vibration modes are calculated considering a coupling effect between the PCV shell and suppression pool water using conical shell finite elements and ring fluid elements. The same expanding functions and free vibration modes are used to express strain energy, kinetic energy and work done by inertia forces due to input acceleration considering locally varying thickness of the PCV and attached masses. The equations of motion are obtained by substituting the strain energy, kinetic energy, and work done by inertia forces into the Lagrange's equation. Calculated natural frequencies, free vibration modes and frequency response functions are compared with numerical results obtained by a general shell finite element analysis and with test results using the large-scale high-performance vibration table of Tadotsu Engineering Laboratory, Nuclear Power Engineering Test Center (NUPEC). The comparison shows a reasonable agreement, leading to better understanding of the dynamic characteristics of the BWR PCV

  20. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

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

  1. 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)

  2. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

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

  3. Outdoor Operational Stability of Indium-Free Flexible Polymer Solar Modules Over 1 Year Studied in India, Holland, and Denmark

    DEFF Research Database (Denmark)

    Angmo, Dechan; Sommeling, Paul M.; Gupta, Ritu;

    2014-01-01

    We present an outdoor interlaboratory stability study of fully printed and coated indium-tin-oxide (ITO)-free polymer solar cell modules in JNCASR Bangalore (India), ECN (Holland), and DTU (Denmark) carried over more than 1 year. The modules comprising a fully printed and coated stack (Ag grid....../PEDOT:PSS/ZnO/P3HT:PCBM/PEDOT:PSS/Ag grid) were prepared in two successive generations and evaluated for outdoor operational stability according to the test protocols laid out by the International Summit on OPV stability (ISOS-3). The modules (70–100 cm2 active area) were encapsulated between two sheets of low...

  4. Stabilization

    Directory of Open Access Journals (Sweden)

    Muhammad H. Al-Malack

    2016-07-01

    Full Text Available Fuel oil flyash (FFA produced in power and water desalination plants firing crude oils in the Kingdom of Saudi Arabia is being disposed in landfills, which increases the burden on the environment, therefore, FFA utilization must be encouraged. In the current research, the effect of adding FFA on the engineering properties of two indigenous soils, namely sand and marl, was investigated. FFA was added at concentrations of 5%, 10% and 15% to both soils with and without the addition of Portland cement. Mixtures of the stabilized soils were thoroughly evaluated using compaction, California Bearing Ratio (CBR, unconfined compressive strength (USC and durability tests. Results of these tests indicated that stabilized sand mixtures could not attain the ACI strength requirements. However, marl was found to satisfy the ACI strength requirement when only 5% of FFA was added together with 5% of cement. When the FFA was increased to 10% and 15%, the mixture’s strength was found to decrease to values below the ACI requirements. Results of the Toxicity Characteristics Leaching Procedure (TCLP, which was performed on samples that passed the ACI requirements, indicated that FFA must be cautiously used in soil stabilization.

  5. Charging of capacitors with double switch. The principle of operation of auto-zero and chopper-stabilized DC amplifiers

    CERN Document Server

    Yordanov, Vasil G; Manolev, Stojan G; Mishonov, Todor M

    2015-01-01

    The principle of operation of auto-zero and chopper-stabilized DC amplifiers, which is realized in many contemporary operational amplifiers is illustrated by a simple experimental setup given at the Open Experimental Physics Olympiad 2014 - "The Day of the Capacitor", held in Sofia and Gevgelija. The Olympiad was organized by the Sofia Branch of the Union of Physicists in Bulgaria and the Regional Society of Physicists of Strumica, Macedonia. In addition to the solution of the secondary school task in the paper is given a detailed engineering description of the patent by Edwin Goldberg and Jules Lehmann, Stabilized direct current amplifier, U.S. Patent 2,684,999 (1949).

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

    International Nuclear Information System (INIS)

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

  7. Thermal aging evaluation of casting stainless steel under BWR environment

    International Nuclear Information System (INIS)

    Effect of thermal aging under BWR condition on material properties of casting stainless steel were evaluated by such as Charpy impact test, using replaced BWR component material. Solution heat treatment was performed to the same material and the material properties were obtained. Comparing each material test results, impact value of thermal aging material was lower than solution heat treatment material. By the results, thermal aging effect on material properties under BWR condition was confirmed. The material properties were compared with model equation using PLM evaluation and conservativeness of model equation was confirmed. (author)

  8. Use of the TRAC/BF1 code in BWR reactors instability studies

    International Nuclear Information System (INIS)

    The RETRAN and TRAC codes are examples of temporary codes that are used to analyze the stability of B.W.R. Although, in many cases, this codes present good results and predict the expected behaviour, they are very sensitive to the variations of core modeling, like for example, variations in the number of cells. This can question seriously the reliability and obviously the acceptability of the analysis done with this temporary codes. In this paper we present a work using the TRAC-BF1 code to simulate the in-phase and out-of-phase oscillations, and the influence of the chose of some parameters. (author)

  9. Product Evaluation Task Force Phase Two report for BWR/PWR dissolver wastes

    International Nuclear Information System (INIS)

    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)

  10. Erythropoietin improves operant conditioning and stability of cognitive performance in mice

    Directory of Open Access Journals (Sweden)

    Ehrenreich Hannelore

    2009-07-01

    Full Text Available Abstract Background Executive functions, learning and attention are imperative facets of cognitive performance, affected in many neuropsychiatric disorders. Recently, we have shown that recombinant human erythropoietin improves cognitive functions in patients with chronic schizophrenia, and that it leads in healthy mice to enhanced hippocampal long-term potentiation, an electrophysiological correlate of learning and memory. To create an experimental basis for further mechanistic insight into erythropoietin-modulated cognitive processes, we employed the Five Choice Serial Reaction Time Task. This procedure allows the study of the effects of erythropoietin on discrete processes of learning and attention in a sequential fashion. Results Male mice were treated for 3 weeks with erythropoietin (5,000 IU/kg versus placebo intraperitoneally every other day, beginning at postnatal day 28. After termination of treatment, mice were started on the Five Choice Serial Reaction Time Task, with daily training and testing extending to about 3 months. Overall, a significantly higher proportion of erythropoietin-treated mice finished the task, that is, reached the criteria of adequately reacting to a 1.0 sec flash light out of five arbitrarily appearing choices. During acquisition of this capability, that is, over almost all sequential training phases, learning readouts (magazine training, operant and discriminant learning, stability of performance were superior in erythropoietin-treated versus control mice. Conclusion Early erythropoietin treatment leads to lasting improvement of cognitive performance in healthy mice. This finding should be exploited in novel treatment strategies for brain diseases.

  11. Quantitative evaluations of BWR regional oscillations using higher harmonics subcriticalities

    International Nuclear Information System (INIS)

    A quantitative study of a mechanism for BWR regional stability has been carried out from the higher harmonics viewpoint using a three-dimensional higher harmonics analysis code and is reported in this paper. The results show that the first azimuthal harmonics subcriticality takes a relatively small value under the regionally unstable condition. The calculated patterns of the first azimuthal harmonics agree quite well with the measured regional oscillation mode patterns. Comparing the subcriticality and the steady state power distribution, it is shown that the distribution exists whose first azimuthal harmonics subcriticality takes a small value. This implies that regionally unstable conditions can be estimated from fundamental mode distributions. The decomposition method of the oscillated power responses into the harmonics modes is presented. The results show that core-wide oscillation power responses almost entirely consist of the fundamental mode and regional oscillation power responses almost entirely consist of the first azimuthal harmonics mode. This indicates that the regional oscillation is the phenomenon in which the first azimuthal harmonics mode oscillates on the basis of the fundamental mode. (author)

  12. The design and use of proficiency based BWR reactor maintenance and refuelling training mockups

    International Nuclear Information System (INIS)

    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

  13. Reduced moderation BWR with advanced recycle system (BARS)

    International Nuclear Information System (INIS)

    lattice. Glass rods coated with transparent electric resistant heater material are adopted in this experiment. The two-phase (steam and water) flow behavior in the test channel has been recorded by video camera and analyzed. These experiments have been done under the atmospheric pressure in order to make it easy to observe. The second one is establishing of the BT correlation equation on tight lattice bundle. Mini tight lattice bundle experiments were planned to study the correlation. These experiment include 7-rods hexagonal bundle experiment and 14-rods square bundle experiment. Various rod configurations on tight lattice have been examined in 7-rods bundle experiments for the parametric study on tight lattice. The 14-rods bundle experiment aims the measurement of square channel effect. One of the important purposes of these experiments is to verify the existing the BT correlation equation such as Arai's equation on tight lattice bundle. These experiments have been done in Toshiba's high temperature and high-pressure BWR experimental facility for stability and transient test (BEST). The third one is cooling property for tight lattice in hypothetical accident evaluation. Counter-current-flow-limitation (CCFL) experiment was planned for this purpose. (author)

  14. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  15. 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)

  16. Impurity hideout/hideout return at the Susquehanna 2 BWR

    International Nuclear Information System (INIS)

    An impurity hideout return study was performed at the Susquehanna 2 BWR to provide an understanding of impurity hideout processes during normal operation and their impact on high temperature solution chemistry in corrosion product deposits on the fuel. Limited hideout return data obtained during shutdowns at 10 BWRs previously had indicated reasonable consistency with expectations based on MULTEQ high temperature solution chemistry modeling of hideout processes. Observations at Susquehanna 2 were consistent with expectations. Cumulative returns of species forming precipitates at low concentration factors above the bulk water concentration, e.g., calcium, magnesium, sulfate and silica were much greater than those of species having a minimal tendency to precipitate, e.g., sodium and chloride. Solutions present in the fuel cladding surface during normal operation were predicted to contain high concentrations (0.1 to 2 molal) of sodium, potassium, chloride, sulfate, silica and nitrate. The predicted solution pH at 300 degrees C was 9.4 (neutral pH = 5.5). The increase in conductivity observed during and after shutdown was shown to be due to solubilization of precipitates with retrograde solubilities rather than chemical/resin intrusion. Variations in reactor water concentrations during reactor water cleanup system isolation and power reductions were consistent with predictions developed from a mass balance around the reactor coolant system

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

    International Nuclear Information System (INIS)

    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)

  18. Small scale BWR core debris eutectics formation and melting experiment

    International Nuclear Information System (INIS)

    A small scale experiment has recently been performed at Oak Ridge under the auspices of the BWR Severe Accident Technology (BWRSAT) program to provide information concerning the formation of mixtures during heatup of representative BWR reactor vessel bottom head debris and to determine the composition and melting temperatures of these mixtures. The initial structure of the bottom head debris layers modeled in the experiment was taken from the results of recent BWR Accident Response (BWRSAR) code predictions for the short-term station blackout accident sequence. The experimental results provide useful information concerning the mixtures formed and their proportions and properties. The observed run-off of a stainless steel-zirconium eutectic alloy supports the contention that the initial pour from a BWR reactor vessel would consists of molten metals at relatively low temperatures. (orig.)

  19. Improvement of BWR radioactive waste disposal system

    International Nuclear Information System (INIS)

    Description is made here about the improvement of the disposal systems for liquid and solid radioactive waste in recent BWR plants. Regarding the improvement of liquid waste disposal systems in Toshiba, emphasis was laid on crud removal and laundry drain treatment; as a result, a crud removal system utilizing a centrifugal separation mechanism and an evaporative enriching system using anti-foaming agent have been practicalized. An important problem left for solution is the method of volume reduction of solid waste. A method attracting attention of late is the plastic solidification which is far superior to the conventional cement and bitumen solidification. In Toshiba, this method is promising to be practicalized, in which the waste is dried and then solidified with thermosetting resin. (author)

  20. Reactor fueling of BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable the pattern exchange for control rods during burning in Control Cell Core type BWR reactors. Constitution: A plurality of control cells are divided into a plurality of groups such that the control cells is aparted from each other by way of at least two fuel assemblies other than the control cells with respect to the vertical and lateral directions of the reactor core cross section, as well as they are in adjacent with control cells of other groups with respect to the orthogonal direction. This enables to perform the pattern exchange for the control rods during burning in the control cell core with ease, and the control blade and the story effect harmful to the mechanical soundness of fuels can thus be suppressed. (Moriyama, K.)

  1. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    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)

  2. TRU transmutation type BWR fuel assembly

    International Nuclear Information System (INIS)

    The BWR fuel assembly is formed by bundling a plurality of fuel rods and a water channel disposed at the center of the assembly by a plurality of spacers. An upper tie plate and a lower tie plate are disposed to upper and lower portions of the fuel rods and the water channel respectively. An upper end plug of the water channel is attached detachably to a cylindrical main body of the water channel. A zircaloy tube incorporating TRU nuclides is contained and secured in the water channel. The zircaloy tube has such a structure as capable of incorporating and sealing oxides or metal materials containing TRU nuclides. Since the zircaloy tube containing TRU nuclides is contained not in fuel region but in the water rod, the loaded uranium amount of fuels is not reduced but the reactivity can be ensured. (I.N.)

  3. BWR startup and shutdown activity transport control

    International Nuclear Information System (INIS)

    This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 oF (

  4. Development of a numerical code for the analysis of the linear stability of the U1 and U2 reactors of the CNLV

    International Nuclear Information System (INIS)

    The computer code ANESLI-1 developed by the CNSNS and UAM-I, has the main goal of making stability analysis of nuclear reactors of the BWR type, more specifically, the reactors of the U1 and U2 of the CNLV. However it can be used for another kind of applications. Its capacity of real time simulator, allows the prediction of operational transients, and conditions of dynamic steady states. ANESLI-1 was developed under a modular scheme, which allows to extend or/and to improve its scope. The lineal stability analysis predicts the instabilities produced by the wave density phenomenon. (Author)

  5. In-reactor ECP measurements in BWR plants

    International Nuclear Information System (INIS)

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

  6. Operation of the counter-rotating type pump-turbine unit installed in the power stabilizing system

    International Nuclear Information System (INIS)

    This serial research intends to put a unique power stabilization system with a pumped storage into practical use. The pumped storage is equipped with a counter-rotating type pump-turbine unit whose operating mode can be shifted instantaneously in response to the fluctuation of power from renewable resources. This paper verifies that the system is reasonably effective to stabilize the fluctuating power. It is necessary to quickly increase the rotational speed when the operation is shifted from the turbine to the pumping modes, because the unit cannot pump-up water from a lower reservoir at a slow rotational speed while keeping gross/geodetic head constant. The maximum hydraulic efficiency at the turbine mode is close to the efficiency of the counter-rotating type hydroelectric unit designed exclusively for the turbine mode. The system is also provided for a pilot plant to be operated in the field

  7. Optimizing stability, transport, and divertor operation through plasma shaping for steady-state scenario development in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, C T; Ferron, J R; Luce, T C; Petrie, T W; Politzer, P A; Rhodes, T L; Doyle, E J; Makowski, M A; Kessel, C; DeBoo, J C; Groebner, R J; Osborne, T H; Snyder, P B; Greenfield, C M; La Haye, R J; Murakami, M; Hyatt, A W; Challis, C; Prater, R; Jackson, G L; Park, J; Reimerdes, H; Turnbull, A D; McKee, G R; Shafer, M W; Groth, M; Porter, G D; West, W P

    2008-12-19

    Recent studies on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] have elucidated key aspects of the dependence of stability, confinement, and density control on the plasma magnetic configuration, leading to the demonstration of nearly noninductive operation for >1 s with pressure 30% above the ideal no-wall stability limit. Achieving fully noninductive tokamak operation requires high pressure, good confinement, and density control through divertor pumping. Plasma geometry affects all of these. Ideal magnetohydrodynamics modeling of external kink stability suggests that it may be optimized by adjusting the shape parameter known as squareness ({zeta}). Optimizing kink stability leads to an increase in the maximum stable pressure. Experiments confirm that stability varies strongly with {zeta}, in agreement with the modeling. Optimization of kink stability via {zeta} is concurrent with an increase in the H-mode edge pressure pedestal stability. Global energy confinement is optimized at the lowest {zeta} tested, with increased pedestal pressure and lower core transport. Adjusting the magnetic divertor balance about a double-null configuration optimizes density control for improved noninductive auxiliary current drive. The best density control is obtained with a slight imbalance toward the divertor opposite the ion grad(B) drift direction, consistent with modeling of these effects. These optimizations have been combined to achieve noninductive current fractions near unity for over 1 s with normalized pressure of 3.5<{beta}{sub N}<3.9, bootstrap current fraction of >65%, and a normalized confinement factor of H{sub 98(y,2)}{approx}1.5.

  8. Design study of Thorium-232 and Protactinium-231 based fuel for long life BWR

    Science.gov (United States)

    Trianti, N.; Su'ud, Z.; Riyana, E. S.

    2012-06-01

    A preliminary design study for the utilization of thorium added with 231Pa based fuel on BWR type reactor has been performed. In the previous research utilization of fuel based Thorium-232 and Uranium-233 show 10 years operation time with maximum excess-reactivity about 4.075% dk/k. To increase reactor operation time and reduce excess-reactivity below 1% dk/k, Protactinium (Pa-231) is used as Burnable Poison. Protactinium-231 has very interesting neutronic properties, which enable the core to reduce initial excess-reactivity and simultaneously increase production of 233U to 231Pa in burn-up process. Optimizations of the content of 231Pa in the core enables the BWR core to sustain long period of operation time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation ratio we can get reactor core with longer operation time, 20 ˜ 30 years operation without fuel shuffling or refuelling, with average power densities maximum of about 35 watt/cc, and maximum excess-reactivity 0.56% dk/k.

  9. Reliable Operation for 14500 h of a Wavelength-Stabilized Diode Laser System on a Microoptical Bench at 671 nm

    DEFF Research Database (Denmark)

    Sumpf, Bernd; Maiwald, Martin; Müller, André;

    2012-01-01

    Reliability tests for wavelength-stabilized compact diode laser systems emitting at 671 nm are presented. The devices were mounted on microoptical benches with the dimensions of 13 mm $\\times\\,$4 mm. Reflecting Bragg gratings were used for wavelength stabilization and emission width narrowing. The...... reliability tests were performed at 25$^{\\circ}{\\rm C}$ and at an output power up to 10 mW per micrometer stripe width of the gain medium. Reliable operation could be demonstrated over a test time up to 14500 h at an output power up to 1.0 W. Environmental tests using random vibrations with acceleration up to...

  10. Japanese BWR utilities joint research on fatigue monitoring system of RPV

    International Nuclear Information System (INIS)

    Considering the plant life extension (PLEX) of nuclear power plants (NPPs) in the future, Japanese BWR electric utilities are carrying out the research project on the fatigue monitoring system based on the actual plant operating data. Reactor Pressure Vessel (RPV) is one of the most critical components of BWR when utilities think about PLEX. RPV is designed by the detail stress analysis using the method such as Finite Element Analysis (FEA). Design stress analysis is performed in accordance with design basis thermal cycles for design life of 40 years and it sometimes reports considerably high fatigue usage factors at such portions of RPV as stud bolt, feedwater nozzle and support skirt. However, because of the conservatism on the design analysis and in view of operational experiences, actual service life of NPPs is expected to be much longer than the design life. For PLEX, it is important to understand the excessive conservatism in the design analysis and the realistic fatigue usage factor (UF) based on the actual plant operating data in order to assess the remaining life time of RPV. To get the realistic UF, the stress analysis requires not only the actual operating process data but also the realistic thermal boundary conditions such as heat transfer coefficient. And stress analysis method must be simple in order to compute a vast number of stresses during plant life time in a short time. Therefore, Japanese BWR utilities jointly with BWR manufacturers established programs (1) to acquire plant operating data on-line for specific parameters used in stress analysis, (2) to evaluate margins in the design using measured plant data and best estimate boundary conditions for stress analysis, and (3) to establish simplified stress analysis method for fatigue evaluation. This paper describes the achievements of this research and the research activities in the future. In the program (1), on-line plant data acquisition system, named OPEDAS, has been developed and installed

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

    International Nuclear Information System (INIS)

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

  12. Water injection system for turbine driven BWR type reactor

    International Nuclear Information System (INIS)

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

  13. A damage tolerance analysis of an old BWR pressure vessel

    International Nuclear Information System (INIS)

    A program has been completed in Sweden where the oldest of the Swedish BWR-units has been subjected to an extensive investigation regarding signs of aging and degradation. The core and all the internal parts were removed and the stripped pressure vessel was decontaminated. Almost every weld in the vessel including all nozzles in the bottom head have then been inspected by aid of UT-inspection. An important part of the programs was the damage tolerance analysis. It has involved postulated surface cracks and embedded cracks (positioned along and across the welds) in all the inspected welds. By using the R6-method the maximum acceptable and critical crack size have then been determined for the most limiting load case and accounting for the individual material properties of each weld, cladding and base material. Of special interest is the core region. The base material is made of A 302 Grade B with rather high Cu- and Ni-contents, which have caused irradiation embrittlement in the beltline region. This implies that during certain cold loading cases, the critical crack size for a postulated surface crack in the core region, can be quite small. However, for load cases during normal operation, the material is on the upper shelf region and the critical crack sizes are large

  14. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

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

  15. 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)

  16. 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)

  17. Effect of Algal Biofilm and Operational Conditions on Nitrogen Removal in Wastewater Stabilization Ponds

    OpenAIRE

    Babu, M.

    2011-01-01

    Discharge of nutrient rich wastewater causes eutrophication of surface water; therefore wastewater treatment before discharge is required. Wastewater stabilization ponds are low cost technology used by developing countries but not effective in nitrogen removal due to low nitrifier biomass in the water column. Introduction of surface area for attachment of nitrifiers has therefore been proposed. This thesis reports the performance of pilot scale wastewater stabilization ponds fitted with baffl...

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

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Gonzalez, Jesus S. [Univ. Politecnica de Madrid (Spain); Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William BJ J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

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

  19. Survey of BWR plant personnel on shutdown safety practices and risk management needs. EPRI Outage Risk Assessment and Management (ORAM) program. Final report, March 1992

    International Nuclear Information System (INIS)

    In response to increasing attention to potential risks during shutdown conditions, nuclear utilities are explicitly addressing safety in their outage planning and management. As part of an integrated package to assist utilities, a survey of BWR utilities was conducted. Detailed interviews with three BWR utilities were supplemented by insights from NRC and BWR Owners Group surveys. The primary objective was to acquire information for use in developing shutdown risk management guidance. Summaries of the interviews and other surveys resulted in thirteen detailed conclusions related to 1) current levels of safety reviews that have a bearing on shutdown operations, 2) the current process for outage planning and management, 3) outage management policies and guidelines that are in use, 4) perceived problems and challenges, and 5) recommendations for future risk management guidance. (author)

  20. Investigation of the load change behaviour of PWR- and BWR fuel rods at positive power ramps

    International Nuclear Information System (INIS)

    The following irradiation experiments have been performed to determine the operational behaviour of fuel rods in LWR during power ramps: a) power ramp experiment in the nuclear power plant of Obrigheim (KWO) with 6 PWR test fuel rods at a burnup of about 14 MWd/kgU. No fuel rod defects have been found. b) preirradiation of 45 segmented fuel rods in KWO and of 8 segmented fuel rods in the reactor of Wuergassen; the preirradiated segments will be ramped at HFR Petten. c) power ramp experiments at HBWR with 8 BWR test fuel rods at burnups of 4-14 MWd/kgU; ramping caused no defects. (orig.)

  1. Effect of SCC growth behavior estimation methods on inspection frequency of BWR primary recirculation loop pipe

    International Nuclear Information System (INIS)

    The current inspection rule for the BWR primary recirculation loop pipes makes some of their weld joints inspected with in five operating years. If seems that one of the backgrounds of the number is the estimation based on the SCC growth behavior analysis. It is well-known that the latest knowledge or information should be input for such analysis to appropriately revise the way of the inspection. In this paper, the effect of the calculation methods for the stress intensity factor on the remaining life of the 300A riser pipe weld joint which currently provides the critical life time. (author)

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

    International Nuclear Information System (INIS)

    A dynamic model of a nuclear power plant, including a boiling water reactor, high- and low-pressure turbines, moisture separator, reheater, condenser, feedwater heaters and feedwater pump, was developed. The model is one-dimensional except for the nuclear part of the reactor, which is based on the point kinetics equation, and the condenser model and feedwater pump model. It has been used to study different transients occuring during normal operating conditions and for evaluating the control systems of a BWR nuclear power plant. Particular emphasis was laid on the reactor pressure control system and the recirculation flow control system. (author)

  3. 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

  4. Investigation of stability of multi free surfaces at transient operation for fast breeder demonstration reactors in Japan

    International Nuclear Information System (INIS)

    The Japanese demonstration fast breeder reactor (JDFBR) is composed of a reactor vessel, intermediate heat exchangers and pump vessels. Every component has a free surface of sodium. Transient operation of the pumps may cause variations of the sodium levels. For the stability of the multiple surfaces, a 1/15 scale model of the JDFBR with 4 loops with a 1000 MWe output power was made to experimentally investigate the stability of 9 free surfaces. In addition, we have developed a computer code to calculate it. The results of the experiments and the calculations agree well with each other. The computer code was successfully verified. The cover gas has an important role to suppress the vibrations of the free surfaces in transient conditions. The sodium level of the JDFBR is stable in all operating conditions, even beyond the design base conditions. (author)

  5. Corrosion product transport and radiation field buildup modelling in the BWR primary system

    International Nuclear Information System (INIS)

    The experimental data obtained from several operating BWRs with a variety of water quality and operating history have been evaluated. The shutdown radiation fields measured at the recirculation piping were found initially to increase over a wide range of buildup rates, but they generally level off in about 5 years of operation. The radiation levels in some plants remain in the vicinity of 100 mR/h during their history. Based on the observed data and known chemical/physical phenomena, major corrosion product transport mechanisms have been hypothesized and a mathematical model of corrosion product transport and radiation field buildup in the BWR primary system has been developed. Calculations using the model are capable of reproducing the observed data well within the experimental uncertainties and confirm that the radiation fields in the operating plants can be controlled and reduced by plant operational practices. (author)

  6. Effect of operation parameters on the flux stabilization of gravity-driven membrane (GDM) filtration system for decentralized water supply.

    Science.gov (United States)

    Tang, Xiaobin; Ding, An; Qu, Fangshu; Jia, Ruibao; Chang, Haiqing; Cheng, Xiaoxiang; Liu, Bin; Li, Guibai; Liang, Heng

    2016-08-01

    A pilot-scale gravity-driven membrane (GDM) filtration system under low gravitational pressure without any pre-treatment, backwash, flushing, or chemical cleaning was carried out to investigate the effect of operation parameters (including operation pressure, aeration mode, and intermittent filtration) on the effluent quality and permeability development. The results revealed that GDM system exhibited an efficient performance for the removal of suspended substances and organic compounds. The stabilization of flux occurred and the average values of stable flux were 6.6, 8.1, and 8.6 Lm(-2) h(-1) for pressures of 65, 120, and 200 mbar, respectively. In contrast, flux stabilization was not observed under continuous and intermittent aeration conditions. However, aeration (especially continuous aeration) was effective to improve flux and alleviate membrane fouling during 1-month operation. Moreover, intermittent filtration would influence the stabilization of permeate flux, resulting in a higher stable flux (ranging from 6 to 13 Lm(-2) h(-1)). The stable flux significantly improved with the increase of intermittent period. Additionally, GDM systems exhibited an efficient recovery of flux after simple physical cleaning and the analyses of resistance reversibility demonstrated that most of the total resistance was hydraulic reversible resistance (50-75 %). Therefore, it is expected that the results of this study can develop strategies to increase membrane permeability and reduce energy consumption in GDM systems for decentralized water supply. PMID:27189452

  7. BWR blowdown/emergency core cooling integral program

    International Nuclear Information System (INIS)

    The Program plan identifying the phased approach to testing has been completed and the first test phase is well underway. Test results to date show the expected lower peak cladding temperatures and slower system blowdown response for a BWR/6 LOCA simulation compared to the BWR/4. Counter-current flow limiting (CCFL) tests show that exact geometric replication was not necessary to simulate CCFL characteristics for a BWR prototype fuel bundle upper tie plate. Separate effects blowdown tests demonstrate the importance of flow length scaling in simulating break geometry for limiting the critical or blowdown flow rate. A new bundle thermal hydraulic method, MAYU04, has been completed. This method is shown to provide a substantial improvement in the prediction of bundle temperatures. The BD/ECC Program represents an important contribution to BWR safety research. Early results from the program have already provided a better understanding of the governing phenomena during hypothetical LOCA simulation tests. Future tests are expected to provide a basis for further improvements in BWR LOCA phenomena modeling

  8. Improving the Operational Stability of PBDTTTz-4 Polymer Solar Cells Modules by Electrode Modification

    DEFF Research Database (Denmark)

    Roth, Bérenger; Benatto, Gisele Alves dos Reis; Corazza, Michael;

    2016-01-01

    PBDTTTz-4 is employed in the ambient manufacturing of fully Roll-to-Roll organic solar cell modules. Modules are manufactured using a novel silver nanowire electrode or a previously reported carbon electrode. The average PCE of carbon modules (3.07%) and AgNW modules (1.46%) shows that PBDTTTz-4 is...... a good candidate for upscaling. Stability measurements following the ISOS standards are used to compare the lifetime of the different modules. In all tests but one, the carbon modules are less stable. The higher stability of AgNW is attributed to the removal of the PEDOT:PSS in the front electrode...

  9. Analysis for steam separators in BWR

    International Nuclear Information System (INIS)

    Steam seperators in BWR are thin walled pipes with special cyclons for seperating steam and water arranged in a hexagonal scheme and welded to the core cover head. The pipes are connected to each other in two planes by a gridwork of small upright sheets also arranged in a hexagonal scheme. At the level of these gridworks there are also shroud rings for holding long screws. The sheets of the gridwork run concentric to the tubes and are interrupted by the tubes. The interiour forces of the gridwork-members are transferred through the tubewalls by circumferential bending which causes extreme elastic ovalizing of the pipe corss-section near to the gridwork connection. The height of these members may be constant or broader at the pipe connection (fig. 1d) for reducing the bending stresses in the pipe walls. If this ovalizing is involved in the elastic behaviour of the gridwork it works like a plate with a Poisson ratio ν approx. equal to 1. Such structures are extremely stiff against equal stresses in perpendicular directions and extremely weak for opposite stresses in both directions. Instead of this interrupted grid work the upper one may be a continues one, arranged tangentially to the tupes. The elastic behaviour of this gridwork is quite different to that of the interrupted one. (orig./GL)

  10. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A fuel assembly of an BWR type reactor of present invention, in which a plurality of fuel rods are arranged in a regular square lattice like configuration include vibration-filled fuel rods in which granular nuclear fuel materials and granular non-nuclear fuel materials having a smaller neutron absorbing cross sectional area are mixed and filled. With such a constitution, the content of the mixed and filled non-nuclear fuel materials in the vibration filled fuel rods is at least 20% by a volume ratio in average in fuel assemblies. In addition, a burnable poison is optionally added and mixed to the granular mixture of the nuclear fuel material and the diluting granules. With such a constitution, the manufacturing cost can be reduced, and the combustion rate of the nuclear fission materials is increased to improve reactor core characteristics, thereby enabling to obtain sufficient Pu loading amount per assembly, and fuel assemblies excellent in flexibility in design and economic property can be obtained. (T.M.)

  11. A BWR pump suction-line 200% break test at ROSA-III Program (RUN 903)

    International Nuclear Information System (INIS)

    The Rig-of-Safety Assessment (ROSA)-III Program conducted a 200% recirculation pump-suction line break test, RUN 903, simulating a loss-of-coolant accident (LOCA) in a boiling water reactor (BWR). In this test, the main recirculation pumps (MRPs) were continuously operated during the transient following the break to study the influence of increased core flow rate on the system responses. This report describes major thermal-hydraulic phenomena observed in this test and presents all the experiment data. The effects of prolonged pump operation on system responses are described in comparison with the results of standard 200% break test, RUN 926, in which the core flow coasted down after the break faster than a scaled BWR LOCA condition. It is shown that the significant core heatup observed during the early blow-down phase in RUN 926 was not observed in RUN 903 due to an additional mass transport (approximately 6% of the initial system mass) from the downcomer into the core shroud as a result of the prolonged pump operation. It is clear that the lower-than-scaled transient core flow rate in the ROSA-III tests significantly affected the core thermal conditions especially during the early blowdown phase. (author)

  12. Laguna Verde nuclear power plant: an experience to consider in advanced BWR design

    International Nuclear Information System (INIS)

    Laguna Verde is a BWR 5 containment Mark II. Designed by GE, two external re-circulation loops, each of them having two speed re-circulation pump and a flow control valve to define the drive flow and consequently the total core flow an power control by total core flow. Laguna Verde Design and operational experience has shown some insights to be considering in design for advanced BRW reactors in order to improve the potential of nuclear power plants. NSSS and Balance of plant design, codes used to perform nuclear core design, margins derived from engineering judgment, at the time Laguna Verde designed and constructed had conducted to have a plant with an operational license, generating with a very good performance and availability. Nevertheless, some design characteristics and operational experience have shown that potential improvements or areas of opportunity shall be focused in the advanced BWR design. Computer codes used to design the nuclear core have been evolved relatively fast. The computers are faster and powerful than those used during the design process, also instrumentation and control are becoming part of this amazing technical evolution in the industry. The Laguna Verde experience is the subject to share in this paper. (author)

  13. Stability Improvements in a Rail-to-Rail Input/Output, Constant Gm Operational Amplifier, at 0.4 V Operation, Using the Low-Voltage DTMOS Technique

    OpenAIRE

    Correia, Joana; Mancelos, Nuno; Goes, João

    2014-01-01

    The use of the dynamic threshold MOS (DTMOS) technique is evaluated in a two-stage rail-to-rail Input/Output, constant Gm amplifier. The proper choice of specific transistors in which the technique should be used is presented, as well as the resulting improvements, mainly regarding stability of the circuit at low voltage operation. The DTMOS technique is used in the NMOS transistors of the folded-cascode input stage, allowing the circuit to be stable at VDD = 0.4 V, with equivalent gain and g...

  14. An overview of the BWR ECCS strainer blockage issues

    Energy Technology Data Exchange (ETDEWEB)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-03-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, {open_quotes}Containment Emergency Sump Performance,{close_quotes} and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts.

  15. Application of expert system and neural network in diagnosis during BWR ATWS sequences

    International Nuclear Information System (INIS)

    A prototype operator aid system employing an expert system and neural network is designed to help the plant operator during a BWR ATWS accident. The expert system is the driver of the inference engine, it consists of IF -- THEN -- and DO -- format rules developed from the knowledge base. A back propagation neural network is used when the operator can not supply the needed information to the expert system. Data of various plant parameters are fed into a pretrained neural network for transient identification. The case signature is then fed into the expert system, where a decision is made regarding the proper operator response. Testing results show that the neural network can retrieve the transients correctly even when random noise is added or the input data is incomplete. The computer simulation of the integrated system has also been demonstrated

  16. Structural response of DN15-tubes under radiolysis gas detonation loads for BWR safety applications

    International Nuclear Information System (INIS)

    A U-shaped DN15 tube with 15 mm ID, 3 mm wall thickness was exposed to radiolysis gas (2H2+O2) detonation loads to investigate the structural stability of typical BWR tubes. Radiolysis gas at ambient temperatures was used at initial pressure up to 70 bar. The effect of transient detonation loads with peak pressures up to 1540 bar on the tube response was studied with strain gauges and simultaneous local pressure measurements. The strain measurements demonstrated that the tube material remained in the elastic response regime for initial radiolysis gas pressures of up to 20 bar. For the case with 30 and 70 bar initial pressure, local plastic deformations were observed under peak detonation pressures of 540 and 1540 bar, respectively. The measured strain values could be well explained with a simplified analysis of the elastic-plastic material behaviour under quasi-static loading conditions. Based on the measured strain data for the DN-15 tube, upper and lower bounds were estimated for the burst pressures of the failed pipes in the Brunsbuettel and the Hamaoka-1 NPP events. The experiments provide new data for the validation of structural dynamic codes and models of the response of typical BWR tubes under radiolysis gas detonation loads. (authors)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

  19. Anterior transarticular screw fixation as a conventional operation for rigid stabilization

    OpenAIRE

    Manabu Sasaki; Katsumi Matsumoto; Koichiro Tsuruzono; Kazuhiro Yoshimura; Katsuhiko Shibano; Kazuo Yonenobu

    2014-01-01

    Background: Anterior transarticular screw (ATS) fixation is a useful surgical option for atlantoaxial (AA) stabilization. This report presents a revised ATS method for AA fusion. Methods: A 79-year-old male presented with AA instability attributed both to an old odontoid fracture and severe degeneration of the lateral atlantoaxial joints (LAAJs). ATS fixation was performed through the conventional anterior cervical approach. The longest screw trajectories were planned preoperatively using...

  20. Remote operation of the vertical plasma stabilization @ the GOLEM tokamak for the plasma physics education

    Czech Academy of Sciences Publication Activity Database

    Svoboda, V.; Kocman, J.; Grover, O.; Krbec, Jaroslav; Stöckel, Jan

    96-97, October (2015), s. 974-979. ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] Institutional support: RVO:61389021 Keywords : tokamak technology * remote participation * plasma stabilization Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.152, year: 2014 http://dx.doi.org/10.1016/j.fusengdes.2015.06.044

  1. Ulam stability for fractional differential equations in the sense of Caputo operator

    Directory of Open Access Journals (Sweden)

    Rabha W. Ibrahim

    2012-12-01

    Full Text Available In this paper, we consider the Hyers-Ulam stability for the following fractional differential equations, in the sense ofcomplex Caputo fractional derivative defined, in the unit disk: cDßzf(z=G(f(z, cDázf(z,zf‘(z;z 0<á<1<ß<2 . Furthermore,a generalization of the admissible functions in complex Banach spaces is imposed and applications are illustrated.

  2. Influence of design parameters and operating conditions on the performance of stabilization ponds in the treatment of swine waste

    Directory of Open Access Journals (Sweden)

    Flávio Lúcio da Silva

    2009-12-01

    Full Text Available This study aims to evaluate the performance the system of stabilization ponds in full scale, operated in series, treating swine manure. The system was constituted of: one anaerobic pond, two facultative ponds and one maturation pond. The research was carried out for 350 days, in swine production manure treatment Plant, in Sao Jose of Lapa/MG. The following parameters were analyzed: temperature, pH, biochemical oxygen demand (BOD, chemical oxygen demand (COD, total solids (TS and volatile total solids (VTS according to the methods established by the Standard Methods for Examination of Water and Wastewater (1998. The facultative ponds operating with low hydraulic retention times (7 e 7,7 days and with high superficial organic loading rate the BOD (1288 kgBOD ha-1 d-1 e 640 kgBOD ha-1 d-1, affecting its performance. The system stabilization ponds showed, in terms of BOD, COD, TS and VTS mean removal efficiency of 86, 86, 85 and 88%, respectively.Keys-words: swine manure, ponds stabilization, treatment system.

  3. Cobra-IE Evaluation by Simulation of the NUPEC BWR Full-Size Fine-Mesh Bundle Test (BFBT)

    Energy Technology Data Exchange (ETDEWEB)

    Burns, C. J. and Aumiler, D. L.

    2006-04-26

    The COBRA-IE computer code is a thermal-hydraulic subchannel analysis program capable of simulating phenomena present in both PWRs and BWRs. As part of ongoing COBRA-IE assessment efforts, the code has been evaluated against experimental data from the NUPEC BWR Full-Size Fine-Mesh Bundle Tests (BFBT). The BFBT experiments utilized an 8 x 8 rod bundle to simulate BWR operating conditions and power profiles, providing an excellent database for investigation of the capabilities of the code. Benchmarks performed included steady-state and transient void distribution, single-phase and two-phase pressure drop, and steady-state and transient critical power measurements. COBRA-IE effectively captured the trends seen in the experimental data with acceptable prediction error. Future sensitivity studies are planned to investigate the effects of enabling and/or modifying optional code models dealing with void drift, turbulent mixing, rewetting, and CHF.

  4. Cobra-IE Evaluation by Simulation of the NUPEC BWR Full-Size Fine-Mesh Bundle Test (BFBT)

    International Nuclear Information System (INIS)

    The COBRA-IE computer code is a thermal-hydraulic subchannel analysis program capable of simulating phenomena present in both PWRs and BWRs. As part of ongoing COBRA-IE assessment efforts, the code has been evaluated against experimental data from the NUPEC BWR Full-Size Fine-Mesh Bundle Tests (BFBT). The BFBT experiments utilized an 8 x 8 rod bundle to simulate BWR operating conditions and power profiles, providing an excellent database for investigation of the capabilities of the code. Benchmarks performed included steady-state and transient void distribution, single-phase and two-phase pressure drop, and steady-state and transient critical power measurements. COBRA-IE effectively captured the trends seen in the experimental data with acceptable prediction error. Future sensitivity studies are planned to investigate the effects of enabling and/or modifying optional code models dealing with void drift, turbulent mixing, rewetting, and CHF

  5. A proposal for accident management optimization based on the study of accident sequence analysis for a BWR

    International Nuclear Information System (INIS)

    The paper describes a proposal for accident management optimization based on the study of accident sequence and source term analyses for a BWR. In Japan, accident management measures are to be implemented in all LWRs by the year 2000 in accordance with the recommendation of the regulatory organization and based on the PSAs carried out by the utilities. Source terms were evaluated by the Japan Atomic Energy Research Institute (JAERI) with the THALES code for all BWR sequences in which loss of decay heat removal resulted in the largest release. Identification of the priority and importance of accident management measures was carried out for the sequences with larger risk contributions. Considerations for optimizing emergency operation guides are believed to be essential for risk reduction. (author)

  6. Stabilities of transients in networks with wind energy. Modelization for operation

    International Nuclear Information System (INIS)

    Two main issues are addressed in this paper. First, models for dynamic studies will be described, justifying simplifications that cannot be performed for other studies, such as power quality studies. Variable speed wind turbines (with doubly fed induction generator), very widely used nowadays, require also an adequate modelling of the control system, according to the time constants and integration step that are used in transient stability studies. Secondly, transient stability studies involving wind energy conversion systems are presented. These studies will include fixed speed and variable speed devices, and several situations on the grid are considered. The results shall be compared and conclusions are drawn out from them. The problem of the required protections is addressed, in relation with the requirements of connection standards for wind energy conversion systems. These subjects will be applied to the Spanish situation, where a great amount of wind penetration is foreseen for the next years. Index terms: Wind power generation, power system stability, doubly fed induction generation. (Author)

  7. Examination of dissimilar metal welds in BWR and PWR piping

    International Nuclear Information System (INIS)

    This paper addresses dissimilar metal weld examinations at PWRS. Surveys were conducted to document the dissimilar metal weld configurations at PWR plants and to update the information known about dissimilar metal weld configurations at BWR plants. The experiences which BWR utilities have had with dissimilar metal weld examinations are documented and include: correct identification of IGSCC, indications thought to be IGSCC but were actually fabrication flaws, and difficulties encountered with the examination of dissimilar metal welds after stress improvement. An experimental program was conducted which verified that the longitudinal wave procedures developed for BWRs are also applicable to PWR designs

  8. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

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

  9. Corrosion potential monitoring and its simulation in BWR conditions

    International Nuclear Information System (INIS)

    A simulation algorithm of corrosion potentials for BWR plant materials has been demonstrated. Cathodic and anodic electrochemical kinetic equations have been derived by analyzing kinetic models involving water radiolysis products of oxygen, hydrogen peroxide, and hydrogen. Corrosion rates of type 304 stainless steel with respect to corrosion potentials are formulated by numerical analysis as well. An electrochemical mixed potential theorem is applied to compute corrosion potentials. Flow rate effects of coolants on corrosion potentials of plant structural materials are expressed as a function of diffusion layer thickness. A fundamental technique and a theory to simulate corrosion potentials have been developed. Corrosion potentials in BWR conditions can be simulated by these results

  10. Low-flow stability tests at Peach Bottom Atomic Power Station unit 2 during cycle 3

    International Nuclear Information System (INIS)

    This report under RP1020-2 is a description of the results from the low-flow stability tests conducted at the Peach Bottom-2 boiling water reactor (BWR) in 1978 during Cycle 3. These tests were a continuation of the series begun in 1977 under RP1020-1. Results from the earlier tests have been documented in EPRI Topical Report NP-564. As expected, the Peach Bottom-2 BWR core is very stable even at minimum core flow rate conditions. These data constitute high-quality proof of some long-held hypotheses about BWR stability at high power-to-flow ratios. In the future, these data will become part of the qualification data base for the improved BWR stability predictive computer codes being developed and tested under RP1384

  11. Operational stability of solution based zinc tin oxide/SiO2 thin film transistors under gate bias stress

    Directory of Open Access Journals (Sweden)

    Asal Kiazadeh

    2015-06-01

    Full Text Available In this study, we report solution-processed amorphous zinc tin oxide transistors exhibiting high operational stability under positive gate bias stress, translated by a recoverable threshold voltage shift of about 20% of total applied stress voltage. Under vacuum condition, the threshold voltage shift saturates showing that the gate-bias stress is limited by trap exhaustion or balance between trap filling and emptying mechanism. In ambient atmosphere, the threshold voltage shift no longer saturates, stability is degraded and the recovering process is impeded. We suggest that the trapping time during the stress and detrapping time in recovering are affected by oxygen adsorption/desorption processes. The time constants extracted from stretched exponential fitting curves are ≈106 s and 105 s in vacuum and air, respectively.

  12. Effect of two impurities and zinc on stress corrosion cracking of stainless steel and nickel alloys in BWR environments

    International Nuclear Information System (INIS)

    Boiling water reactors (BWRs) operate with very high purity water with only small additions of dissolved hydrogen and, most recently, noble metals. However, even operation with very low conductivity water (e.g., 0.07 μS/cm) coolant will not prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel and nickel alloys under atypical oxygenated conditions. The presence of certain impurities dissolved in the coolant can dramatically increase the propensity of this most insidious form of environmentally-assisted cracking. The goal of this paper is to present the effect of effect of chloride and sulfate plus zinc on the IGSCC propensities of BWR piping and reactor internals under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions. While it is well documented the sulfate and chloride are particularly aggressive in promoting IGSCC of BWR structural materials, several anions such as chromate and nitrate have little impact while of zinc added as zinc oxide appears to be beneficial. To emphasize the effect of impurities on the structural integrity of BWR components in perspective, the BWR fleet's most severe documented water chemistry transient, where the conductivity reached on 232 μS/cm with 21.2 ppm chloride and 93.8 ppm sulfate, will be presented. For example, on-line real-time crack growth rate measurements using the highly accurate reversing DC potential drop technique revealed a crack growth rate increase by almost a factor of 300 for an Alloy 182 weld metal compact tension fracture mechanics specimen during this raw water transient. The recommendations for subsequent plant inspection and start up after this transient will also be discussed where the value of real time crack growth rate monitoring cannot be overemphasized. (author)

  13. Attention and its role in the operator's work. [control stability in man machine systems

    Science.gov (United States)

    Shvetsov, O. P.

    1975-01-01

    A complex attentiometer investigation of the distribution, redirection, and concentration of attention during an operator's work notes the following stages: (1) General attentiveness is still not adequately expressed in the beginning; and (2) operator self-control of actions develops and gradually decreases errors in redirecting and distributing attention. A definite relationship is found between the improvement of concentration, distribution and redirection of attention and automation of sensorimotor performance. Excercises prove less effective in redirection of attention.

  14. Passivation of gas microstrip detectors and stability of long-term operation

    International Nuclear Information System (INIS)

    We have studied the long-term operation of gas microstrip detectors which have been passivated with a layer of nickel oxide. We have used as the active gas CF4/isobutane (80 : 20) and three different types of substrates: Tedlar, glass and Upilex. In all three cases we found that the detectors are stable after passivation and can operate for a month without changes in gain at rates of MHz. The total accumulated charge was approximately 100 mC. ((orig.))

  15. Effect of operational variables on nitrogen transformations in duckweed stabilization ponds

    OpenAIRE

    Caicedo Bejarano, J.R.

    2005-01-01

    There is a diversity of conventional technologies available for removal of pollutants from wastewater. Most of these technologies are aerobic alternatives with high construction cost and high energy consumption and require skilled personal for operation and maintenance. As a consequence, only countries with a high gross national product (GNP) can afford these options. Where these technologies were introduced in developing countries, in most cases these could not be operated sustainably, leadi...

  16. Teleoperator comfort and psychometric stability: Criteria for limiting master-controller forces of operation and feedback during telemanipulation

    Science.gov (United States)

    Wiker, Steven F.; Hershkowitz, Elaine; Zik, John

    1989-01-01

    The following question is addressed: How much force should operators exert, or experience, when operating a telemanipulator master-controller for sustained periods without encountering significant fatigue and discomfort, and without loss of stability in psychometric perception of force. The need to minimize exertion demands to avoid fatigue is diametrically opposed by the need to present a wide range of force stimuli to enhance perception of applied or reflected forces. For 104 minutes subjects repetitiously performed a series of 15 s isometric pinch grasps; controlled at 5, 15, and 25 percent of their maximum voluntary strength. Cyclic pinch grasps were separated by rest intervals of 7.5 and 15 s. Upon completion of every 10 minute period, subjects interrupted grasping activities to gage the intensity of fatigue and discomfort in the hand and forearm using a cross-modal matching technique. A series of psychometric tests were then conducted to determine accuracy and stability in the subject's perception of force experienced. Results showed that onset of sensations of discomfort and fatigue were dependent upon the magnitude of grasp force, work/rest ratio, and progression of task. Declines in force magnitude estimation slopes, indicating a reduction in force perception sensitivity, occurred with increased grasp force when work/rest ratios were greater than 1.0. Specific recommendations for avoiding discomfort and shifts in force perception, by limiting pinch grasp force required for master-controller operation and range of force reflection or work/rest ratios, are provided.

  17. Optimal Stabilization of Social Welfare under Small Variation of Operating Condition with Bifurcation Analysis

    Science.gov (United States)

    Chanda, Sandip; De, Abhinandan

    2015-07-01

    A social welfare optimization technique has been proposed in this paper with a developed state space based model and bifurcation analysis to offer substantial stability margin even in most inadvertent states of power system networks. The restoration of the power market dynamic price equilibrium has been negotiated in this paper, by forming Jacobian of the sensitivity matrix to regulate the state variables for the standardization of the quality of solution in worst possible contingencies of the network and even with co-option of intermittent renewable energy sources. The model has been tested in IEEE 30 bus system and illustrious particle swarm optimization has assisted the fusion of the proposed model and methodology.

  18. NONCONFORMING STABILIZED FINITE ELEMENT METHODS BASED ON RIESZ-REPRESENTING OPERATORS

    Institute of Scientific and Technical Information of China (English)

    DuanHuoyuan

    1999-01-01

    Following the framework of the finite element methods based on Riesz-representingoperators developed by Duan Huoyuan in 1997,through discrete Rieszonsome virtual(non-) conforming finite-dimensional subspaces,a stabilization formulation is presented for the Stokes problem by employing nonconforming elements. This formulation is uni-tormly coercive and not subject to the Babus Ka-Brezzi condition,and the resulted linearalgebraic system is positive definite with the spectral condition number O(h-2).Quasi-optimal error bounds are obtained,which is consistent with the interpolation properties of the finite elements used.

  19. Timing criteria for supplemental BWR emergency response equipment

    International Nuclear Information System (INIS)

    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)

  20. Anterior transarticular screw fixation as a conventional operation for rigid stabilization

    Directory of Open Access Journals (Sweden)

    Manabu Sasaki

    2014-01-01

    Full Text Available Background: Anterior transarticular screw (ATS fixation is a useful surgical option for atlantoaxial (AA stabilization. This report presents a revised ATS method for AA fusion. Methods: A 79-year-old male presented with AA instability attributed both to an old odontoid fracture and severe degeneration of the lateral atlantoaxial joints (LAAJs. ATS fixation was performed through the conventional anterior cervical approach. The longest screw trajectories were planned preoperatively using multiplanar reconstruction computed tomography (CT scans, with entry points of the screws situated at the midpoint on the inferior border of the axial body. The surgical exposure was limited to opening at the entry points alone. Our retractor of choice was the Cusco speculum; it sufficiently secured space for utilizing the required instruments for screw placement while offering sufficient protection of soft tissues. Cannulated full-threaded bicortical screws stabilized the LAAJs. Screw insertion required a significant amount of coronal angulation up to the superior articular process of the atlas under open-mouth and lateral fluoroscopy image guidance. After ATS fixation, bone grafting was performed between the posterior laminae of the axis and the atlas through a conventional posterior approach. Results: Bony fusion between the atlas and the axis was confirmed radiographically. Arthrodesis of the LAAJs occurred despite no bone grafting. Conclusions: Rigid fixation of the LAAJs was obtained by our ATS technique, indicating that it is an alternative method for AA fixation when posterior rigid internal fixation is not applicable.

  1. Study of transient rod extraction failure without RBM in a BWR

    International Nuclear Information System (INIS)

    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)

  2. Derivation of the field equations and stability parameters for three operating modes of the three-dimensional quadrupole

    International Nuclear Information System (INIS)

    Derivations of the field equations and stability parameters of a quadrupole ion trap, operating in different modes, are given in detail and discussed. The modes considered are: I. Conventional, analogous to the mass filter: a combination voltage, phi0, is applied to the ring electrode and -phi0 to the end caps. II. Simple: phi0 is applied to the ring and the end-caps are earthed. III. A new mode in which the dc component of phi0 is applied to the end-caps and the rf component to the ring. Possible applications of mode III to studies of ion-molecule reaction pathways are also given. (Auth.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  5. 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)

  6. Stability monitoring of the Dodewaard boiling-water reactor

    International Nuclear Information System (INIS)

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

  7. Annual meeting on nuclear technology 1980. Technical meeting: Operating experiences

    International Nuclear Information System (INIS)

    In addition to general experiences, experiences in reactor operation with relation to the Phenix reactor, KNK-2 reactor, the AVR reactor, the BWR-type KKI-reactor, the Philippsburg-1 reactor and the Obrigheim reactor are described. (DG)

  8. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2014-10-15

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  9. Effect of operational variables on nitrogen transformations in duckweed stabilization ponds

    NARCIS (Netherlands)

    Caicedo Bejarano, J.R.

    2005-01-01

    There is a diversity of conventional technologies available for removal of pollutants from wastewater. Most of these technologies are aerobic alternatives with high construction cost and high energy consumption and require skilled personal for operation and maintenance.

  10. Pool swell in a nuclear containment wetwell. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, T.

    1976-04-01

    A brief description is presented of scale model tests conducted to study LOCA induced wetwell pool swelling in the BWR Mk 1 containment pressure suppression system. The Mk 1 containment configuration is described together with the scale model design, the conduct of the tests, and the experimental results. (DG)

  11. Age-related degradation of BWR control rod drives

    International Nuclear Information System (INIS)

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

  12. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Redding, J.R. [GE Nuclear Energy, San Jose, CA (United States)

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

  13. Stability of operation versus temperature of a three-phase clock-driven chaotic circuit

    International Nuclear Information System (INIS)

    We evaluate the influence of temperature on the behavior of a three-phase clock-driven metal—oxide—semiconductor (MOS) chaotic circuit. The chaotic circuit consists of two nonlinear functions, a level shifter, and three sample and hold blocks. It is necessary to analyze a CMOS-based chaotic circuit with respect to variation in temperature for stability because the circuit is sensitive to the behavior of the circuit design parameters. The temperature dependence of the proposed chaotic circuit is investigated via the simulation program with integrated circuit emphasis (SPICE) using 0.6-μm CMOS process technology with a 5-V power supply and a 20-kHz clock frequency. The simulation results demonstrate the effects of temperature on the chaotic dynamics of the proposed chaotic circuit. The time series, frequency spectra, bifurcation phenomena, and Lyapunov exponent results are provided. (general)

  14. Selection of design and operational parameters in spindle-holder-tool assemblies for maximum chatter stability by using a new analytical model

    OpenAIRE

    Ertürk, Alper; Erturk, Alper; Budak, Erhan; Özgüven, H. Nevzat; Ozguven, H. Nevzat

    2007-01-01

    In this paper, using the analytical model developed by the authors, the effects of certain system design and operational parameters on the tool point FRF, thus on the chatter stability are studied. Important conclusions are derived regarding the selection of the system parameters at the stage of machine tool design and during a practical application in order to increase chatter stability. It is demonstrated that the stability diagram for an application can be modified in a predictable manner ...

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

    International Nuclear Information System (INIS)

    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

  16. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors

    International Nuclear Information System (INIS)

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  17. On Stability of Open-Loop Operation without Rotor Information for Brushless DC Motors

    OpenAIRE

    Zhong Wu; Haotun Lyu; Yongli Shi; Di Shi

    2014-01-01

    Open-loop operation mode is often used to control the Brushless DC Motors (BLDCMs) without rotor position sensors when the back electromotive force (EMF) is too weak due to the very low rotor velocity. The rotor position information is not necessary in this mode and the stator windings are supplied with voltages under a certain ratio of the amplitude to the frequency. However, the rotor synchronization will be destroyed once if the commutation instant is inappropriate. In order to improve the...

  18. Critical stability conditions in skimmed milk crossflow microfiltration: impact on operating modes

    OpenAIRE

    Gésan-Guiziou, Geneviève; Daufin, Georges; Boyaval, Evelyne

    2000-01-01

    There exists a critical ratio (convection towards the membrane/erosion) in crossflow microfiltration, MF, below which there is no marked fouling by colloidal particles and above which performance are altered: sharp increase of fouling, reduced operating time, large decrease in permeability and solute transmission. This paper outlines the impact of the critical ratio on skimmed milk MF processing (separation of casein micelles from the soluble proteins) and gives objective elements (taking int...

  19. Endurance and stability of some surface meteorological sensors under land- and ship-based operating environments

    Digital Repository Service at National Institute of Oceanography (India)

    Mehra, P.; Desai, R.G.P.; Joseph, A.; VijayKumar, K.; Dabholkar, N.; Prabhudesai, S.; Nagvekar, S.; Agarvadekar, Y.

    Tema (Ghana, West Africa). Ship based systems have been operational from research vessels Sagar Kanya, Sagar Sampada, and Sagar Sukti. In view of the importance of data quality in scientific studies, it is vital to keep track of the performance... of instruments from which data have been collected. Several researchers have reported the development of meteorological systems and the performance of meteorological sensors incorporated therein (e.g., [2 to 11]). However, field-performances of such sensors...

  20. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  1. Stability Analysis of Pitch-regulated, Variable Speed Wind Turbines in Closed Loop Operation Using a Linear Eigenvalue Approach

    International Nuclear Information System (INIS)

    A multi-body aeroelastic design code based on the implementation of the combined aeroelastic beam element is extended to cover closed loop operation conditions of wind turbines. The equations of a controller for variable generator speed and pitch controlled operation in high wind speeds are combined with the aeroelastic equations of motion for the complete wind turbine, in order to provide a compound servo-aeroelastic system of equations. The control equations comprise linear differential equations for the pitch and generator torque actuators, the control feedback elements (PI control) and the various filters acting on the feedback signals. In its non-linear form the dynamic equations are integrated in time to provide the reference state, while upon linearization of the system and transformation in the non-rotating frame, the linear stability equations are derived. Stability results for a multi-MW wind turbine, show that the coupling of the controller dynamics with the aeroelastic dynamics of the machine is important and must be taken into account in view of defining the controller parameters

  2. The effect of metal-buffer bilayer drain/source electrodes on the operational stability of the organic field effect transistors

    International Nuclear Information System (INIS)

    In this paper, we have investigated experimentally the effect of different drain/source (D/S) electrodes and charge injection buffer layers on the electrical properties and operational stability of a stilbene organic field effect transistor (OFET). The results show that the organic buffer layer of copper phthalocyanine (CuPc) considerably improves the electrical properties of the transistors, but has a negligible effect on their temporal behavior. On the other hand, inorganic metal-oxide buffer layer of molybdenum oxide (MoO3) drastically changes both the electrical properties and operational stability. The functionalities of this metal-oxide tightly depend on the properties of the D/S metallic electrodes. OFETs with Al/MoO3 as the bilayer D/S electrodes have the best electrical properties: field effect mobility μeff = 0.32 cm2 V−1 s−1 and threshold voltage VTH = − 5 V and the transistors with Ag/MoO3 have the longest operational stability. It was concluded that the chemical stability of the metal/metal-oxide or metal/organic interfaces of the bilayer D/S electrodes determine the operational stability of the OFETs. - Highlights: • The effect of buffer layers on the performance of the stilbene OFETs has been investigated. • Inorganic buffer layer improved the electrical and temporal behaviors simultaneously. • Organic buffer layer only changes the electrical properties. • Chemical stability of the interfaces determines the operational stability of the transistor

  3. Full MOX BWR core physics experiment. Experimental and analysis results of 9x9 reference core

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) has been conducting an experimental program that is aimed to obtain a comprehensive data base for validation of core analysis methods applied to the full MOX ABWR and also for high burn up MOX fuel expected in the future. As a part of this program, JNES has been performing a MOX core physics experimental program, FUBILA, with collaboration of a French Consortium (CEA and COGEMA). The experiments has been designed to obtain the core physics data of the operating conditions of the full MOX BWR cores consisting of high burn up BWR MOX assemblies. The experiments started from January 2005 and completed in September 2006 at the EOLE critical facility of the CEA Cadarache center in France. Theoretical analysis of the experimental data has been also carried out with a continuous energy Monte Carlo calculation and a deterministic method with major nuclear data libraries. This report presents the outline of the FUBILA program, the measured data of the critical mass and the power distribution of a 9x9 Ref core, the first experimental core of the seven cores of FUBILA, and the core analysis by the continuous energy Monte Carlo calculation code MVP. (author)

  4. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO2 fuel in Japan. The design concept should be compatible with UO2 fuel design. High burnup UO2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO2 8 x 8 array fuel developed for a second step of UO2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  5. Application of EASY5 and MMS modules to BWR controller design

    International Nuclear Information System (INIS)

    The application of EPRI's MMS Library and BCS' EASY5 simulation language to the design of a digital feedwater control system for the Monticello Boiling Water Nuclear Power Plant is discussed. In order to first design and then verify the digital feedwater controller algorithms, a digital simulation model of the Monticello plant was constructed using a combination of custom designed modules, existing MMS two-phase library modules, and standard modules available in the EASY5 library. Details of the process models, namely the BWR nuclear steam supply system, the steamline piping, and the feedwater piping are described in a companion paper. Details of the models for the existing BWR turbine pressure inlet pressure control and recirculation flow control system are described. These models are required to be operational during the transient analysis portion of the feedwater controller design verification, since they interact strongly with the reactor steam flow and water level. The design of the digital feedwater flow control loop is described. Its design is of particular interest because it requires consideration of control loop interaction and is, therefore, a simple example of multivariable non-interacting control design

  6. Example of a practical demonstration of in-plant inspection methodology (BWR pipework); the demonstration process

    International Nuclear Information System (INIS)

    Following the discovery of widespread cracking in the recirculation piping system at the Nine Mile Point Unit 1 plant, the US Nuclear Regulatory Commission (NRC) made the requirement that the operating BWR plants demonstrate the effectiveness of their ultrasonic testing procedures in detecting intergranular stress corrosion cracks (IGSCC) in large diameter recirculation system pipes. These requirements were issued in the form of bulletins 'IEB 82-03' in October of 1982 and 'IEB 83-02' in March of 1983. EPRI, through the DPRI NDE Centre, took the lead in establishing a program so that the 24 affected BWR plants (nine addressed by IEB 82-03 and 15 addressed by IEB 83-02) could satisfy the NRC requirements with minimum impact on plant schedules. This meant that not only the preparation activities had to be conducted rapidly, but the demonstration process for each team could not last longer than a day, which in itself restricted the size of the test set. This paper describes the elements of the demonstration process and presents a summary of the performances of the inspection teams who took part in the IEB 82-03 and IEB 83-02 activities. (author)

  7. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Cheng L. Y.; Baek J.; Cuadra,A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic to initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.

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

    International Nuclear Information System (INIS)

    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)

  9. Analysis of containment venting following a core damage at a BWR Mark I using THALES-2

    International Nuclear Information System (INIS)

    Analysis of containment venting following a core damage at a boiling water reactor (BWR) Mark I using THALES-2 was performed. In this analysis, the effect of various parameters, namely, the areas of the vent path, containment venting pressure, and accident sequences on the containment thermodynamic response, and radionuclide transport and release in the containment venting at a BWR was examined. The code THALES-2B developed by the Japan Atomic Energy Research Institute (JAERI) was used in this analysis. The model plant in this analysis was the Browns Ferry plant. From this analysis was found that the 4-inch pipe of containment venting flow path is sufficient to maintain the containment pressure in the specified range if the containment was pressurized by the decay heat power. The entrainment by the pool swelling as well as by the flashing was not occurred during the containment venting. The source terms are not sensitive to the variation of containment venting flow path area. The containment venting pressure operation setting point has important rule in the containment venting. In the containment venting, the source terms are not sensitive to the accident sequence, except for Sr source term. In order to get better understanding on the containment venting strategy, the following analyses are necessary. Analyses of accident sequence which has a high power such as anticipated transient without scram are necessary, as well as analyses of accident sequence which pressurize the containment before the core damage. (author)

  10. Strain-induced corrosion cracking in ferritic components of BWR primary circuits

    International Nuclear Information System (INIS)

    The present final report of the RIKORR project is a summary of a literature survey and of the experimental work performed by PSI on the environmentally-assisted cracking (EAC) and dynamic strain ageing (DSA) susceptibility of low-alloy steels (LAS) in high-temperature (HT) water. Within this project, the EAC crack growth behaviour of different low-alloy RPV steels, weld filler and weld heat-affected zone materials has been investigated under simulated transient and steady-state BWR/NWC power operation conditions. The strain-induced corrosion cracking (SICC) / low-frequency corrosion fatigue (CF) and stress corrosion cracking (SCC) crack growth behaviour of different low-alloy RPV steels under simulated transient and stationary BWR/NWC conditions was characterized by slow rising load / low-frequency corrosion fatigue and constant load / periodical partial unloading / ripple load tests with pre-cracked fracture mechanics specimens in oxygenated HT water at temperatures of either 288, 250, 200 or 150 oC. Modern high-temperature water loops, on-line crack growth monitoring and fractographic analysis by scanning electron microscopy (SEM) were used to quantify the cracking response. (author)

  11. Cable Stability

    CERN Document Server

    Bottura, L

    2014-01-01

    Superconductor stability is at the core of the design of any successful cable and magnet application. This chapter reviews the initial understanding of the stability mechanism, and reviews matters of importance for stability such as the nature and magnitude of the perturbation spectrum and the cooling mechanisms. Various stability strategies are studied, providing criteria that depend on the desired design and operating conditions.

  12. Determination of stability of steel girder structures of large surface mining equipment - calculation of the operating strength factor according to the first revision of TGL 13472

    Energy Technology Data Exchange (ETDEWEB)

    Roessiger, R.

    1987-01-01

    The TGL 13472 industrial standard calculation method is explained for steel girder stability and fatigue with the example of a bucket wheel excavator boom. The excavator operates in block mining with the bucket wheel boom slewing 90 degrees. Bucket wheel cutting and girder load forces are determined by a computer program. Their values are listed. Equations for maximum load fatigue and operating strength factor are given along with stability results. Time required for calculations is regarded as reasonable. 4 refs.

  13. Long-term stability of underground operated CZT detectors based on the analysis of intrinsic 113Cd β--decay

    Science.gov (United States)

    Ebert, J.; Gößling, C.; Gehre, D.; Hagner, C.; Heidrich, N.; Klingenberg, R.; Kröninger, K.; Nitsch, C.; Oldorf, C.; Quante, T.; Rajek, S.; Rebber, H.; Rohatsch, K.; Tebrügge, J.; Temminghoff, R.; Theinert, R.; Timm, J.; Wonsak, B.; Zatschler, S.; Zuber, K.

    2016-06-01

    The COBRA collaboration operates a demonstrator setup at the underground facility Laboratori Nazionali del Gran Sasso (LNGS, located in Italy) to prove the technological capabilities of this concept for the search for neutrinoless double beta-decay. The setup consists of 64 (1×1×1) cm3 Cadmium-Zinc-Telluride (CZT) detectors in Coplanar-Grid (CPG) configuration. One purpose of this demonstrator is to test if reliable long-term operation of CZT-CPG detectors in such a setup is possible. The demonstrator has been operated under ultra low-background conditions for more than three years and collected data corresponding to a total exposure of 218 kg days. The presented study focuses on the long-term stability of CZT detectors by analyzing the intrinsic, fourfold forbidden non-unique 113Cd single beta-decay. It can be shown that CZT detectors can be operated stably for long periods of time and that the 113Cd single beta-decay can be used as an internal monitor of the detector performance during the runtime of the experiment.

  14. Long-Term Stability of Underground Operated CZT Detectors Based on the Analysis of Intrinsic $^{113}$Cd \\beta$^{-}$-Decay

    CERN Document Server

    Ebert, J; Gehre, D; Hagner, C; Heidrich, N; Klingenberg, R; Kroeninger, K; Nitsch, C; Oldorf, C; Quante, T; Rajek, S; Rebber, H; Rohatsch, K; Tebruegge, J; Temminghoff, R; Theinert, R; Timm, J; Wonsak, B; Zatschler, S; Zuber, K

    2015-01-01

    The COBRA collaboration operates a demonstrator setup at the underground facility LNGS (Laboratori Nazionali del Gran Sasso, located in Italy) to prove the technological capabilities of this concept for the search for neutrinoless double beta-decay. The setup consists of 64 $(1\\times\\!1\\times\\!1)$ cm$^{3}$ CZT detectors in CPG configuration. One purpose of this demonstrator is to test if reliable long-term operation of CZT-CPG detectors in such a setup is possible. The demonstrator has been operated under ultra low-background conditions since more than three years and collected data corresponding to an exposure of 218 kg$\\cdot$days. The presented study focuses on the long-term stability of CZT detectors by analyzing the intrinsic, fourfold forbidden non-unique $^{113}$Cd single beta-decay. It can be shown that CZT detectors can be operated stably for long periods of time and that the $^{113}$Cd single beta-decay can be used as an internal monitor of the detector performance during the runtime of the experimen...

  15. Development of a numerical code for the analysis of the linear stability of the U1 and U2 reactors of the CNLV; Desarrollo de un codigo numerico para el analisis de estabilidad lineal de los reactores de las U1 y U2 de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa P, G.; Estrada P, C.E. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Nunez C, A.; Amador G, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, Mexico D.F. (Mexico)

    2001-07-01

    The computer code ANESLI-1 developed by the CNSNS and UAM-I, has the main goal of making stability analysis of nuclear reactors of the BWR type, more specifically, the reactors of the U1 and U2 of the CNLV. However it can be used for another kind of applications. Its capacity of real time simulator, allows the prediction of operational transients, and conditions of dynamic steady states. ANESLI-1 was developed under a modular scheme, which allows to extend or/and to improve its scope. The lineal stability analysis predicts the instabilities produced by the wave density phenomenon. (Author)

  16. Step Sizes for Strong Stability Preservation with Downwind-Biased Operators

    KAUST Repository

    Ketcheson, David I.

    2011-08-04

    Strong stability preserving (SSP) integrators for initial value ODEs preserve temporal monotonicity solution properties in arbitrary norms. All existing SSP methods, including implicit methods, either require small step sizes or achieve only first order accuracy. It is possible to achieve more relaxed step size restrictions in the discretization of hyperbolic PDEs through the use of both upwind- and downwind-biased semidiscretizations. We investigate bounds on the maximum SSP step size for methods that include negative coefficients and downwind-biased semi-discretizations. We prove that the downwind SSP coefficient for linear multistep methods of order greater than one is at most equal to two, while the downwind SSP coefficient for explicit Runge–Kutta methods is at most equal to the number of stages of the method. In contrast, the maximal downwind SSP coefficient for second order Runge–Kutta methods is shown to be unbounded. We present a class of such methods with arbitrarily large SSP coefficient and demonstrate that they achieve second order accuracy for large CFL number.

  17. Nonlinear effects in Paul traps operated in the second stability region: analytical analysis and numerical verification.

    Science.gov (United States)

    Xiong, Caiqiao; Zhou, Xiaoyu; Zhang, Ning; Zhan, Lingpeng; Chen, Suming; Nie, Zongxiu

    2014-11-01

    Paul trap working in the second stability region has long been recognized as a possible approach for achieving high-resolution mass spectrometry (MS), which however is still far away from the experimental implementations because of the narrow working area and inefficient ion trapping. Full understanding of the ion motional behavior is helpful for solving the problem. In this article, the ion motion in a superimposed octopole field, which was characterized by the nonlinear Mathieu equation, was solved analytically using Poincare-Lighthill-Kuo (PLK) method. This method equivalently described the nonlinear disturbance by an effective quadrupole field with perturbed Mathieu parameters, a(u) and q(u), which would bring huge convenience in the studies of nonlinear ion dynamics and was, therefore, used for rapid evaluation of the nonlinear effects of ion motion. Fourth-order Runge-Kutta method (4th R-K) indicated the error of PLK for characterizing the frequency shift of ion motion was within 15%. PMID:25183226

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Radiometric calibration and stability of the Landsat-8 Operational Land Imager (OLI)

    Science.gov (United States)

    Markham, Brian L.; Barsi, Julia A.; Kaita, Edward; Ong, Lawrence; Morfitt, Ron A.; Haque, Md. O.

    2015-09-01

    Landsat-8 and its two Earth imaging sensors, the Operational Land Imager (OLI) and Thermal Infrared Sensor (TIRS) have been operating on-orbit for 2 1/2 years. The OLI radiometric calibration, which is monitored using on-board lamps, on-board solar diffusers, the moon and vicarious calibration techniques has been stable to within 1% over this period of time. The Coastal Aerosol band, band 1, shows the largest change at about 1% over the period; all other bands have shown no significant trend. OLI bands 1- 4 show small discontinuities in response (+0.1% to 0.2%) beginning about 7 months after launch and continuing for about 1 month associated with a power cycling of the instrument, though the origin of the recovery is unclear. To date these small changes have not been compensated for, but this will change with a reprocessing campaign that is currently scheduled for Fall 2015. The calibration parameter files (each typically covering a 3 month period) will be updated for these observed gain changes. A fitted response to an adjusted average of the lamps, solar and lunar results will represent the trend, sampled at the rate of one value per CPF.