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Sample records for bwr turbine trip

  1. Peach Bottom Turbine Trip Simulations with RETRAN Using INER/TPC BWR Transient Analysis Method

    International Nuclear Information System (INIS)

    The work described in this paper is benchmark calculations of pressurization transient turbine trip tests performed at the Peach Bottom boiling water reactor (BWR). It is part of an overall effort in providing qualification basis for the INER/TPC BWR transient analysis method developed for the Kuosheng and Chinshan plants. The method primarily utilizes an advanced system thermal hydraulics code, RETRAN02/MOD5, for transient safety analyses. Since pressurization transients would result in a strong coupling effect between core neutronic and system thermal hydraulics responses, the INER/TPC method employs the one-dimensional kinetic model in RETRAN with a cross-section data library generated by the Studsvik-CMS code package for the transient calculations. The Peach Bottom Turbine Trip (PBTT) tests, including TT1, TT2, and TT3, have been successfully performed in the plant and assigned as standards commonly for licensing method qualifications for years. It is an essential requirement for licensing purposes to verify integral capabilities and accuracies of the codes and models of the INER/TPC method in simulating such pressurization transients. Specific Peach Bottom plant models, including both neutronics and thermal hydraulics, are developed using modeling approaches and experiences generally adopted in the INER/TPC method. Important model assumptions in RETRAN for the PBTT test simulations are described in this paper. Simulation calculations are performed with best-estimated initial and boundary conditions obtained from plant test measurements. The calculation results presented in this paper demonstrate that the INER/TPC method is capable of calculating accurately the core and system transient behaviors of the tests. Excellent agreement, both in trends and magnitudes between the RETRAN calculation results and the PBTT measurements, shows reliable qualifications of the codes/users/models involved in the method. The RETRAN calculated peak neutron fluxes of the PBTT

  2. Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

    International Nuclear Information System (INIS)

    Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)

  3. ZZ BWRTT, BWR Turbine Trip Transient Benchmark Based on Peach-Bottom 2

    International Nuclear Information System (INIS)

    Description of program or function: This benchmark project is established to challenge the coupled system T-H/neutron kinetics codes against a Peach-Bottom-2 (a GE-designed BWR/4) turbine trip transient with a sudden closure of the turbine stop valve. Three-turbine trip (TT) transients at different power levels were performed at the Peach Bottom (PB)-2 BWR/4 Nuclear Power Plant (NPP) prior to shutdown for refueling at the end of Cycle 2 in April 1977. The second test is selected for the benchmark problem to investigate the effect of the pressurization transient, (following the sudden closure of the turbine stop valve) on the neutron flux in the reactor core. In a best-estimate manner the test conditions approached the design basis conditions as closely as possible. The actual data were collected, including a compilation of reactor design and operating data for Cycles 1 and 2 of PB and the plant transient experimental data. The transient was selected for benchmark, because it is a dynamically complex event for which neutron kinetics in the core was coupled with thermal-hydraulics in the reactor primary system. The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. The purpose of this proposal is to establish a BWR TT benchmark exercise, based on a

  4. Application of the RELAP5 code for simulation of three turbine trip transients at the Peach Bottom Unit 2 BWR

    International Nuclear Information System (INIS)

    The RELAP5/MOD1, cycle 14 code was applied to simulation of three turbine trip tests conducted on the Peach Bottom Unit 2 boiling water reactor (BWR). The three turbine trip tests analyzed were initiated from different steady-state core power and primary flow conditions, thus providing a useful data set for evaluation of code performance and input modeling techniques. Simulation of these transients with RELAP5 required that the code be modified to include a jet pump model and to improve the RELAP5 steam separator model. The input model was derived from a RETRAN-01 model of the Peach Bottom Unit 2 Plant. Judgment was required in the selection of certain code input parameters such as moderator density and discharge coefficient. The final results obtained with the RELAP5 code were good simulations of all three Peach Bottom Unit 2 turbine trips for the first ten seconds of the transients

  5. Assessment study of the coupled code RELAP5/PARCS against the Peach Bottom BWR turbine trip test

    International Nuclear Information System (INIS)

    The modeling of complex transients in nuclear power plants (NPP) remains a challenging topic for best estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used for simulating transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. In this framework, the Peach Bottom BWR turbine trip experiment 2 is considered. The test involves a rapid positive reactivity addition into the core generated by a water hammer load. To perform a numerical simulation of such phenomenon a reference case was calculated using the coupled code RELAP5/PARCS. An overall data comparison shows good agreement between calculated and measured pressure wave trend in the core region. However, the predicted power response during the excursion phase did not match correctly the experimental tendency. For this purpose, a series of sensitivity analyses have been carried out to identify the most probable reasons of such discrepancy. It was found out that the uncertainties related to the cross-sections modeling and to the thermal-hydraulic closure relationships are the main source of the incorrect power feedback response during the transient

  6. OECD/NRC BWR Turbine Trip Transient Benchmark as a Basis for Comprehensive Qualification and Studying Best-Estimate Coupled Codes

    International Nuclear Information System (INIS)

    An Organisation for Economic Co-operation and Development (OECD)/U.S. Nuclear Regulatory Commission (NRC)-sponsored coupled-code benchmark has been initiated for a boiling water reactor (BWR) turbine trip (TT) transient. Turbine trip transients in a BWR are pressurization events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. In addition, the available real plant experimental data make this benchmark problem very valuable. Over the course of defining and coordinating the BWR TT benchmark, a systematic approach has been established to validate best-estimate coupled codes. This approach employs a multilevel methodology that not only allows for a consistent and comprehensive validation process but also contributes to the study of different numerical and computational aspects of coupled best-estimate simulations. This paper provides an overview of the OECD/NRC BWR TT benchmark activities with emphasis on the discussion of the numerical and computational aspects of the benchmark

  7. Analysis of the peach bottom 2 BWR turbine trip experiment by RELAP 5/3.2 code

    OpenAIRE

    Bousbia-Salah Anis; D'auria Francesko

    2002-01-01

    This paper presents the results of the application of the system of the thermalhydraulic code RELAP5/Mod3.2 in predicting the Peach Bottom Boiling Water Reactor Turbine Trip test. This experiment constitutes a challenge to the capabilities of current computational tools in realistically predicting transient scenarios in nuclear power plants. In fact, it involves strong feedback during the transient between thermalhydraulics and neutronics. In this respect, a reference case was run in order to...

  8. Statistical safety evaluation of BWR turbine trip scenario using coupled neutron kinetics and thermal hydraulics analysis code SKETCH-INS/TRACE5.0

    International Nuclear Information System (INIS)

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method. (author)

  9. Analysis of the peach bottom 2 BWR turbine trip experiment by RELAP 5/3.2 code

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2002-01-01

    Full Text Available This paper presents the results of the application of the system of the thermalhydraulic code RELAP5/Mod3.2 in predicting the Peach Bottom Boiling Water Reactor Turbine Trip test. This experiment constitutes a challenge to the capabilities of current computational tools in realistically predicting transient scenarios in nuclear power plants. In fact, it involves strong feedback during the transient between thermalhydraulics and neutronics. In this respect, a reference case was run in order to simulate the interactions between the generated steam line pressure wave propagation and the instantaneous core void distribution. An overall comparison shows good agreement between the code calculations and the experimental data. A series of sensitivity analyses were also performed in order to assess the code prediction features, as well as to identify uncertainties related to the adopted thermalhydraulic parameters used for the plant modelisation.

  10. Application of the coupled code Athlet-Quabox/Cubbox for the extreme scenarios of the OECD/NRC BWR turbine trip benchmark and its performance on multi-processor computers

    International Nuclear Information System (INIS)

    The OECD/NRC BWR Turbine Trip (TT) Benchmark is investigated to perform code-to-code comparison of coupled codes including a comparison to measured data which are available from turbine trip experiments at Peach Bottom 2. This Benchmark problem for a BWR over-pressure transient represents a challenging application of coupled codes which integrate 3-dimensional neutron kinetics into thermal-hydraulic system codes for best-estimate simulation of plant transients. This transient represents a typical application of coupled codes which are usually performed on powerful workstations using a single CPU. Nowadays, the availability of multi-CPUs is much easier. Indeed, powerful workstations already provide 4 to 8 CPU, computer centers give access to multi-processor systems with numbers of CPUs in the order of 16 up to several 100. Therefore, the performance of the coupled code Athlet-Quabox/Cubbox on multi-processor systems is studied. Different cases of application lead to changing requirements of the code efficiency, because the amount of computer time spent in different parts of the code is varying. This paper presents main results of the coupled code Athlet-Quabox/Cubbox for the extreme scenarios of the BWR TT Benchmark together with evaluations of the code performance on multi-processor computers. (authors)

  11. A turbine trip transient analyses with TRAC-BF1

    International Nuclear Information System (INIS)

    This paper reports a turbine trip transient analysis for a BWR/5 nuclear power plant performed with TRAC-BF1 code, a well-known best-estimate code used for transient analyses of boiling water reactors. The work undertook an evaluation of several safety parameters, such as dome pressure rise, maximum temperature in the fuel, maximum reactor power, total reactivity, and others. 2 refs., 7 figs., 3 tabs

  12. Risk analysis and control measures after turbine trip but reactor not trip

    International Nuclear Information System (INIS)

    Based on the work experience, this paper presents the phenomenon and risks that may occur after turbine trip but reactor not trip while full power operation of PWR nuclear power plant, and put forward control measures to discuss. (authors)

  13. Model with Peach Bottom Turbine trip and thermal-Hydraulic code TRACE V5P3

    International Nuclear Information System (INIS)

    This work is the continuation of the work presented previously in the thirty-ninth meeting annual of the Spanish Nuclear society. The semi-automatic translation of the Thermo-hydraulic model TRAC-BF1 Peach Bottom Turbine Trip to TRACE was presented in such work. This article is intended to validate the model obtained in TRACE, why compare the model results result from the translation with the Benchmark results: NEA/OECD BWR Peach Bottom Turbine Trip (PBTT), in particular is of the extreme scenario 2 of exercise 3, in which there is SCRAM in the reactor. Among other data present in the (transitional) Benchmark , are: total power, axial profile of power, pressure Dome, total reactivity and its components. (Author)

  14. 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)

  15. Analysis of the Peach Bottom 2 turbine trip transients with the plant model ALMOS

    International Nuclear Information System (INIS)

    In April 1977 three turbine trip tests were performed at the Peach Bottom 2 boiling water reactor (BWR) plant, each test starting from a different partial power level and different core flow. A brief description of pertinent portions of the reactor system, the instrumentation, initial conditions and measured transients is presented. The tests were modeled and analysed with the BWR plant model ALMOS. For this purpose the code was implemented with a detailed description of the steam line and bypass system; it is able to account for pressure wave phenomena which are of significance here. Transients obtained from this model are in good agreement with measured counterparts of all tests. This is true especially for the steam dome pressure and the core average power which is coupled closely to the system pressure. Since the ALMOS code offers various options regarding neutronics (point kinetics, 1-D) and two phase slip (HEM, Bankoff) it was possible to investigate their influence on the transient response. In the case of rapid pressure changes as experienced in turbine trips the balance between heat generated within the fuel and heat transferred to the moderator is lost. Therefore, significant errors can arise when the computation of the neutron flux transient with the point kinetics model is based on a void coefficient which is obtained assuming stationary conditions. These errors were quantified by comparison with results of corresponding 1-D kinetics calculations. (orig.)

  16. RETRAN analysis of the turbine trip tests at Peach Bottom Atomic Power Station Unit 2 and at the end of cycle 2. Special report

    International Nuclear Information System (INIS)

    The RETRAN system analysis code is undergoing a comprehensive qualification program by the Electric Power Research Institute and the Utilities' RETRAN Working Group. An important element of this effort is the analysis of the turbine trip tests performed at the Peach Bottom Atomic Power Station Unit 2 boiling water reactor (BWR) during April 1977. The report gives the results obtained from this analysis and from an extensive set of associated sensitivity studies. These results are in good agreement with the measured data, providing important evidence of the code's ability to model BWR turbine trips and similar transients. Other conclusions resulting from this work pertain to needed code improvements and to the sensitivity of the calculated transients to various input parameters. The analysis of the tests and the sensitivity studies performed with RETRAN also provide a better understanding of certain phenomena which are important during BWR transients involving abrupt changes in pressure and flow

  17. Boiling Water Reactor Turbine Trip (TT) Benchmark - Vol. IV Volume IV: Summary Results of Exercise 3

    International Nuclear Information System (INIS)

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling of the core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for that purpose. The present volume is the last in a series of four and summarises the results of the third benchmark exercise, which analyses a turbine trip (TT) in a BWR in its entirety, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. Exercise 3 also analyses four extreme scenarios which allowed participants to test the capabilities of their code(s) in terms of coupling and feedback modelling. The data made available from experiments carried out at the plant make the present benchmark particularly valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4). (authors)

  18. 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.)

  19. RELAP5/MOD1 analysis of Turbine Trip Test 1 at Peach Bottom Unit 2

    International Nuclear Information System (INIS)

    The RELAP5/MOD1, Cycle 14 thermal-hydraulics code was used for analysis of turbine trip transient 1 (TT1) conducted at the Peach Bottom Atomic Power Station Unit 2. The goals were to evaluate the capability of RELAP5/MOD1 for BWR plant transient analysis and to evaluate the adaptability of RETRAN input models for use in RELAP5. The RELAP5/MOD1 input model used in the evaluation was derived from an existing RETRAN input model of Peach Bottom. Alteration of the RETRAN input model was required because of the differences between the codes. Modification of the RELAP5/MOD1, Cycle 14 code was also required to permit effective modeling of jet pumps and the steam separator. The results obtained with the RELAP5/MOD1, Cycle 14 code were in good agreement with the test data. Additional RELAP5 calculations were performed to determine the sensitivity of the results to variations in selected input parameters

  20. Validation of RETRAN-03 by simulating a Peach Bottom turbine trip and boil-off at the FIST facility

    International Nuclear Information System (INIS)

    The RETRAN-03 computer code was validated by simulating two tests that were performed at the Full Integral Simulation Test (FIST) facility. RETRAN-03 results of a Peach Bottom turbine trip (4PTT1) and failure to maintain water level at decay power (T1QUV) are compared with the FIST test data. Sensitivity to various model nodalizations and RETRAN-03 slip options were studied by comparing results of FIST test T1QUV. Core uncovery time is sensitive to the upper downcomer and upper plenum nodalization. The RETRAN-03 analysis of FIST test 4PTT1 was compared to a previous TRAC-BWR analysis of the test

  1. 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)

  2. 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)

  3. Sensitivity of Peach Bottom turbine trip test calculations using RETRAN

    International Nuclear Information System (INIS)

    Fast acting power increase transients are required to test the one-dimensional reactor kinetics used in RETRAN. In addition, methods used to model the steam separator, down comer, steam lines and feedback from turbine control valves, feedwater and recirculation flow control can be tested using the Peach Bottom benchmarks. Gulf States Utilities (GSU) performed simulations of the three Peach Bottom turbine trip tests. These calculations were performed using RETRAN02/MOD4, with kinetics data generated by the GSU's CASMO1/SIMULATE-E/SIMTRAN models. The Peach Bottom tests were held at low power conditions for which reactor vessel pressure and enthalpy distribution data was unavailable. During the computer program benchmark process, it was found that the results showed a large sensitivity to the reactor conditions used to initialize RETRAN. Many other sensitivities were observed such as turbine trip induced pressure wave impact on overall results. Timing and magnitude of the pressure waves were found to be affected by the use of equilibrium or non-equilibrium models, by distribution of separator inertia and by distribution of pressure losses through core and separators. The timing and magnitude of these pressure waves has a significant effect on the peak power and heat flux

  4. Turbine trip with bypass failure analysis of Kuosheng Nuclear Power Plant using TRACE/FRAPCON/FRAPTRAN

    International Nuclear Information System (INIS)

    Kuosheng nuclear power plant (NPP) is located on the northern coast of Taiwan. Its nuclear steam supply system (NSSS) is a type of BWR/6 designed and built by General Electric on a twin unit concept. Each unit includes two loops of recirculation piping and four main steam lines, with the thermal rated power of 2894MWt. Unit 1 will start SPU (Stretch Power Uprate) from Cycle 24 and Unit 2 will start SPU from Cycle 23. The operating power will be 104.7% of the OLTP (Original Licensed Thermal Power). In order to estimate the safety of Kuosheng NPP, the methodology of Kuosheng NPP SPU safety analysis model was developed. There are three main steps considered in this methodology. The first step is the development of the Kuosheng NPP SPU TRACE model. The fuel rods steady state results of FRAPCON were used to input the TRACE model. The next step is the transient analysis of Kuosheng NPP SPU TRACE model. In this paper, the turbine trip without bypass is chosen in order to confirm the maximum pressure of vessel below the acceptance limit of 9.58 MPa. The final step is the fuel rods integrity analysis of FRAPTRAN under the above conditions. The turbine trip without bypass analysis results of TRACE indicate that the Kuosheng NPP SPU TRACE model can predict the behaviors of important parameters and the maximum vessel pressure is below the acceptance limit of 9.58 MPa. Besides, under the above conditions, the results of FRAPCON/FRAPTRAN also depict that the integrity of fuel rods are kept. The maximum of total cladding hoop strain is 0.0016, which is far less than acceptance limit 0.01, indicating that the cladding is safe in this case. And the maximum enthalpy is 52.44 cal/g, which is far less than 170 cal/g specified by the NRC NUREG-0800 standard review plan. (author)

  5. 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)

  6. Calculation of pressure wave inside the steam line with turbine trip

    International Nuclear Information System (INIS)

    After turbine trip, a pressure wave phenomenon happens inside the steam lines and has disadvantageous effects on the steam generator and steam lines. To study this effect, the mathematical models for pressure wave calculation are developed and the calculating results are analyzed

  7. Estimation of skyshine dose from turbine building of BWR plant using Monte Carlo code

    International Nuclear Information System (INIS)

    The Monte Carlo N-Particle transport code (MCNP) was adopted to calculate the skyshine dose from the turbine building of a BWR plant for obtaining precise estimations at the site boundary. In MCNP calculation, the equipment and piping arranged on the operating floor of the turbine building were considered and modeled in detail. The inner and outer walls of the turbine building, the shielding materials around the high-pressure turbine, and the piping connected from the moisture separator to the low-pressure turbine were all considered. A three-step study was conducted to estimate the applicability of MCNP code. The first step is confirming the propriety of calculation models. The atmospheric relief diaphragms, which are installed on top of the low-pressure turbine exhaust hood, are not considered in the calculation model. There was little difference between the skyshine dose distributions that were considered when using and not using the atmospheric relief diaphragms. The calculated dose rates agreed well with the measurements taken around the turbine. The second step is estimating the dose rates on the outer roof surface of the turbine building. This calculation was made to confirm the dose distribution of gamma-rays on the turbine roof before being scattered into the air. The calculated dose rates agreed well with the measured data. The third step is making a final confirmation by comparing the calculations and measurements of skyshine dose rates around the turbine building. The source terms of the main steam system are based on the measured activity data of N-16 and C-15. As a conclusion, we were able to calculate reasonable skyshine dose rates by using MCNP code. (authors)

  8. Technical achievement of advanced 1,100 MW BWR turbine and generator system

    International Nuclear Information System (INIS)

    No.3 plant in Fukushima No.2 Nuclear Power Station, Tokyo Electric Power Co., Inc., started the commercial operation on June 21, 1985, without trouble in the trial operation. In order to meet the needs of improving the economical efficiency of nuclear power generation, many new techniques have been adopted in this plant. In this report, the features of the main new techniques in the turbine-generator system and the results of operation are described. The plant is composed of an improved standard BWR-5 of 3293 MWt and a turbine-generator of rated output 1100 MWe. For improving the operation properties, a 100 % turbine bypass system, a super-rapid response exciting system, new condensate and feed water system, the electro-hydraulic control of feed pump driving turbines, and a low flow rate feed regulating valve were adopted. For improving the reliability, a low pressure rotor of one-body construction and tripled electro-hydraulic control system were adopted. For rationalization, four stages of neck LP feed heaters, a small size turbine building, a flexible turbine-generator foundation, the backwashing for condensers by changing six valves and so on were adopted. Also many improvements of construction method were carried out. (Kako, I.)

  9. Validation of RETRAN-03 by simulating a peach bottom turbine trip and boiloff at the full integral simulation test facility

    International Nuclear Information System (INIS)

    This paper reports that the RETRAN-03 computer code is validated by simulating two tests that were performed at the Full Integral Simulation Test (FIST) facility. The RETRAN-03 results of a turbine trip (test 4PTT1) and failure to maintain water level at decay power (test T1QUV) are compared with the FIST test data. The RETRAN-03 analysis of test 4PTT1 is compared with a previous TRAC-BWR analysis of the test. Sensitivity to various model nodalizations and RETRAN-03 slip options are studied by comparing results of test T1QUV. The predicted thermal-hydraulic responses of both tests agree well with the test data. The pressure response of test 4PTT1 and the boiloff rate for test T1QUV are accurately predicted. Core uncovery time is found to be sensitive to the upper downcomer and upper plenum nodalization. The RETRAN-03 algebraic and dynamic slip options produce similar results for test T1QUV

  10. Transient subchannel analysis of turbine trip without bypass with IC of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Full text: The proposed advanced heavy water reactor (AHWR) is a 750 MWth vertical pressure tube type boiling light water cooled and heavy water moderated reactor. Safety analysis is carried out for various postulated initiating events (PIE) for AHWR. Turbine trip without bypass is one such PIE. Turbine trip results in rise in system pressure. Absence of bypass flow may lead to actuation of relief and safety valves depending upon system characteristics. System pressure rise activates various protective and engineered safeguard systems like reactor trip, isolation condenser and advanced accumulator limiting the consequences of the event. In this paper, a turbine trip without bypass is analysed, using a two-fluid code RELAPS/MOD3.2 [Fletcher, 1995] and subchannel analysis code COBRA IV-I. [Wheeler, 1976] Global analysis does not take into account the effect of different fuel pins with different peaking factors, effect of axial and radial cross flow mixing between adjacent subchannels. Combination of these two codes gives a better insight into the problem. The maximum rated reactor channel which houses 54 pin fuel bundle is modeled for COBRA IV-I simulation. The transient forcing function option of COBRA IV-I, validated by [Iwamura, 1994] with their flow-power transient experiments, has been used for transient thermal hydraulic parameter prediction for this study. Transient inlet pressure, inlet mass flow rate, inlet temperature, outlet enthalpy and the heat flux/power obtained from RELAPS /MOD3.2 simulation are among the boundary conditions employed on the COBRA-IV-I simulation. The thermal-hydraulic parameters predicted by RELAPS/MOD3.2 along with the effect of subchannel flow on the fuel element temperatures predicted by COBRA-IV-I are presented and discussed in this paper

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

    International Nuclear Information System (INIS)

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

  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. 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)

  14. 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

  15. RELAP5 assessment: LOFT turbine trip L6-7/L9-2

    International Nuclear Information System (INIS)

    The RELAP5 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal; hydraulic response of LWRs during accident and off-normal conditions. The RELAP5/MOD1 code is being assessed at SNLA against test data from various integral and separate effects test facilities. As part of this assessment matrix, a turbine trip rapid cooldown transient performed at the LOFT test facility has been analyzed. The results show that RELAP5/MOD1 can predict the experimental behavior of LOFT test L6-7/L9-2 in detail. However, careful selection of modeling options and adjustment of boundary conditions within the experimental uncertainties is required

  16. Turbine coastdown test to evaluate the availability of power supply when reactor trip and LOOP occur at YGN 5

    International Nuclear Information System (INIS)

    The integrity of nuclear fuel is greatly dependent on whether the Reactor Coolant Pumps(RCP) can keep on running during the drop of control rods after a reactor trip signal is generated. It was assumed in the safety analysis for Younggwang Nuclear Power Plant Unit 5 and 6 (YGN 5 and 6) that the RCPs are powered for 3 seconds since the Loss Of Off-site Power(LOOP) will not occur within 3 seconds after reactor and turbine trip simultaneously. A turbine coastdown test was performed at YGN 5 during the power ascension test period to justify the assumption. The purpose of the test was to measure how long the power is supplied to RCPs in virtue of the turbine inertia after a LOOP occurs simultaneously with reactor/turbine trips at 15% reactor power. Based on the data obtained from the test, a computational model was developed and verified. This model is used to simulate the turbine coastdown at various plant power levels including the full power with different magnitudes of houseload. The results of the test and simulation justified the assumption of the delayed interruption of power supply to RCPs

  17. Analysis of the turbine trip transient in Peach Bottom NPP with the coupled codes TRAC-BF1/NEM

    International Nuclear Information System (INIS)

    In the present paper, the results from the simulation of the turbine trip transient Benchmark in Peach Bottom NPP are presented. This simulation has been performed using 0D, 1D and 3D kinetic parameters in the coupled codes TRAC-BF1/NEM. The results of the transient in TRAC-BF1 using the KINPAR methodology and the SIMTAB-1D methodology are compared with the reference SIMULATE3 results. The results of the turbine trip transient with 1D and 3D kinetic parameters are also compared. The power peak value reached in the simulated transient in each of the cases analyzed, using 1D and 3D kinetic parameters, is very different. The reason for that difference is related to the range of the definitions of the cross-sections sets used. The variation of the fuel temperature during the transient is very high, therefore it has to be considered in the cross-section generation

  18. Strengthening the First Line of Defence: Delayed Turbine Trip at SCRAM in Westinghouse type NPP's

    International Nuclear Information System (INIS)

    The availability of Information, Control and Power (ICP) is not treated as a Critical Safety Function (CSF). After the Forsmark (2006) and Fukushima (2011) incidents there is reason to add ICP as a separate CSF. Adding ICP as a separate CSF would possibly lead to procedural adaptations, or even design changes, for Nuclear Power Plants. As an example, this paper focusses on the transitions immediately after a SCRAM. At a SCRAM in many nuclear power plants the turbine is tripped immediately to prevent the extraction of too much heat from the reactor. However this requires a large and fast transition for the entire secondary system. The rescheduled priorities could lead to the wish NOT to trip the turbine before load has been reduced and alternative power has been secured. This paper discusses a 'soft landing' for the turbine by keeping it running after the SCRAM. Turbine control can follow reactor power by controlling the pressure of the available residual steam from the steam generator. With a proper control design this enables a flexible and precise control of primary temperatures without any fast switching in the secondary system during the first 1/2 to 3 minutes. In this period reactor load and turbine power are smoothly lowered to minimum levels during of which automatic preparatory measures can be triggered. The normal transitions can be initiated in a staged form to provide a soft landing for the entire secondary and electrical system. (author)

  19. 三门核电站停机不停堆的运行分析%Operational Analysis of Turbine Trip without Reactor Trip in Sanmen Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    车济尧

    2014-01-01

    三门核电A P1000反应堆在满功率情况下发生汽轮机故障停机事件时,通过快速降功率系统、旁排系统和棒控系统等的快速响应,一回路的参数不会突破安全限值,避免了反应堆停堆,降低了该瞬态对反应堆冷却剂系统的冲击。文章对停机不停堆的实现方式和运行特点进行了详细的分析和阐述,以帮助电站人员对停机不停堆的理解,并提高他们面临瞬态的响应能力。%Reactor is tripped after turbine trip in generate II plant, which prevents the temperature, pressure and water level of reactor coolant system exceeding safety limits and protect the safety of reactor. While Sanmen nuclear power plant is designed to sustain a turbine trip from 100-percent power, without generating a reactor trip, the rapid power reduction system, in conjunction with automatic steam dump control system, and rod control system, is provided to accommodate this abnormal load rejection and to reduce the effects of the transient imposed on the reactor coolant system. In this paper, the design and operation characteristics of turbine trip without reactor trip are analyzed and explained in detail to facilitate the understanding of the concept of turbine trip without reactor trip, and to improve the response ability of plant personnel in the transient.

  20. Pre- and post-test analyses of a KKL turbine trip test at 109% power using RETRAN-3D

    International Nuclear Information System (INIS)

    As part of the PSI/HSK STARS project, pre-test calculations have been performed for a KKL turbine trip test at 109% power using the RETRAN-3D code. In this paper, we first present the results of these calculations, together with a description of the test and a comparison of the results with the measured plant data, and then discuss in more detail the differences between the pre-test results and the plant measurements, including the differences in the initial and boundary conditions, and how these differences influenced the calculated results. Finally, we comment on a series of post-test and sensitivity analyses, which were performed to resolve some of the discrepancies. The results of the pre-test (blind) calculations show good overall agreement with the experimental data, particularly for the maximum in the steam-line mass flow rate following the opening of the turbine bypass valves. This is of critical importance, since the steam-line flow has the least margin to the reactor scram limit. The agreement is especially good since the control rod banks used for the selected rod insertion were changed from those given in the test specification. Following a review of the comparison of the pre-test calculations with the measured data, several deficiencies in the RETRAN-3D model for KKL were identified and corrected as part of the post-test analysis. This allowed for both an improvement in the calculated results, and a deeper understanding of the behaviour of the turbine trip transient. (author)

  1. Assessment of full power turbine trip start-up test for C. Trillo 1 with RELAP5/MOD2. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, M.F.; Moreno, P.; de la Cal, C.; Larrea, E.; Lopez, A.; Santamaria, J.G.; Lopez, E.; Novo, M. [Consejo de Seguridad Nuclear, Madrid (Spain)

    1993-07-01

    C. Trillo I has developed a model of the plant with RELAP5/MOD2/36.04. This model will be validated against a selected set of start-up tests. One of the transients selected to that aim is the turbine trip, which presents very specific characteristics that make it significantly different from the same transient in other PWRs of different design, the main difference being that the reactor is not tripped: a reduction in primary power is carried out instead. Pre-test calculations were done of the Turbine Trip Test and compared against the actual test. Minor problems in the first model, specially in the Control and Limitation Systems, were identified and post-test calculations had been carried out. The results show a good agreement with data for all the compared variables.

  2. Propagation of void fraction uncertainty measures in the RETRAN-3D simulation of the Peach Bottom turbine trip

    Energy Technology Data Exchange (ETDEWEB)

    Vinai, Paolo [Paul Scherrer Institute, Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland); Chalmers University of Technology, Goeteborg (Sweden); Macian-Juan, Rafael [Technische Universitaet Muenchen, Garching (Germany); Chawla, Rakesh [Paul Scherrer Institute, Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland)

    2008-07-01

    The paper describes the propagation of void fraction uncertainty, as quantified by employing a novel methodology developed at PSI, in the RETRAN-3D simulation of the Peach Bottom turbine trip test. Since the transient considered is characterized by a strongly coupling between thermal-hydraulics and neutronics, the accuracy in the void fraction model has a very important influence on the prediction of the power history and, in particular, of the maximum power reached. It has been shown that the objective measures used for the void fraction uncertainty, based on the direct comparison between experimental and predicted values extracted from a database of appropriate separate-effect tests, provides power uncertainty bands that are narrower and more realistic than those based, for example, on expert opinion. The applicability of such an approach to NPP transient best estimate analysis has thus been demonstrated. (authors)

  3. Turbine trip due to inadvertent operation of emergency coolant injection valves during routine testing at Pickering NGS-B

    International Nuclear Information System (INIS)

    During routine emergency coolant injection system (ECIS) testing a control panel operating error resulted in a partial reactor core bypass flow of the primary heat transport coolant, through the odd-numbered D20 injection valves. A reactor trip was averted by prompt operator action, but a turbine generator trip on high boiler level did occur. The unit was successfully recovered from this upset and returned to high power operation. The event was reviewed in some detail, as similar upsets have previously occurred. It was determined that a number of factors contributed to the event, principally in the areas of control panel ergonomics, the test procedure that was used, control room staffing, and the scheduling of safety system testing. A number of corrective measures which have been put in place are discussed, and it is anticipated that these will minimize the risk of re-occurrence of this event in the future. The findings from this investigation are generally applicable to all main control room routine panel testing operations and merit review for their applicability at other facilities. (author). 4 figs

  4. Application of improved air transport data and wall transmission/reflection data in the SKYSINE code to typical BWR turbine skyshine

    International Nuclear Information System (INIS)

    Three basic sets of data, i.e. air transport data and material transmission/reflection data, included in the SKYSHINE program have been improved using up-to-data and methods, and applied to skyshine dose calculations for a typical BWR turbine building. The direct and skyshine dose rates with the original SKYSHINE code show good agreements with MCNP Monte-Carlo calculations except for the distances less than 0.1 km. The results for the improved SKYSHINE code also have agreements with the MCNP code within 10-20%. The discrepancy of 10-20% can be due to the improved concrete transmission data at small incident and exit angles. We still improve the three sets of data and investigate with different calculational models to get more accurate results. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: andyskamx@yahoo.com.mx

    2004-07-01

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

  7. Safety/relief valve quencher loads: evaluation for BWR Mark II and III containments

    International Nuclear Information System (INIS)

    Boiling water reactor (BWR) plants are equipped with safety/relief valves (SRVs) to protect the reactor from overpressurization. Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool produces high-pressure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment inside containment. This report presents the results of the staff's evaluation of SRV loads. The evaluation, however, is limited to the quencher devices used in Mark II and III containments. With respect to Mark I containments, the SRV acceptance criteria are presented in NUREG-0661 issued July 1980. The staff acceptance criteria for SRV loads for Mark II and III containments are presented in this report

  8. 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

  9. 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.

  10. Development and application of a semi-quantitative RCM approach to the reactor and Turbine Closed Cooling Water Systems in BWR

    International Nuclear Information System (INIS)

    Research and development is being performed at Chubu Electric Power Co., Inc. to develop a method for optimization of maintenance. In this study, a Reliability Centered Maintenance (RCM) method is developed that is more quantitative than the traditional function based RCM. This method is being applied to the Reactor (RCCW) and Turbine Closed Cooling Water (TCCW) Systems. The results of this research demonstrate the effectiveness of this approach in judging component 'criticality'. (author)

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

    International Nuclear Information System (INIS)

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

  12. 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)

  13. 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

  14. 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

  15. Results and experiences from the analysis of the OECD PWR MSLB and BWR TT benchmarks by the coupled code system ATHLET-QUABOX/CUBBOX

    International Nuclear Information System (INIS)

    The GRS results of the two QECD/NRC benchmarks for coupled codes - PWR Main Steam Line Break (MSLB) and BWR Turbine Trip (TT) are presented in the paper. Both benchmarks have been calculated for all exercises by the coupled code system ATHLET - QUABOX/CUBBOX developed by GRS. The OECD PWR Main Steam Line Break Benchmark has been defined to validate the coupled code systems with 3D neutronics by comparing solutions of different codes (code-to-code comparison). The BWR TT Benchmark is based on measurements. Both benchmarks proved to be a valuable source for coupled code validation. The sensitivity of results on modelling features is also discussed in the paper. In addition, the effect of different mapping schemes between fuel assemblies of the core loading and the thermal-fluid dynamics on the accuracy of 3D neutronics solutions is shown. The results for the MSLB transient are also evaluated to compare 3D neutronics and point-kinetics solutions in view of integral and local parameters. Thus, the experiences with the coupled code system ATHLET - QUABOX/CUBBOX during the MSLB and TT benchmarks activities are summarised. (author)

  16. 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

  17. 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.)

  18. A subchannel and CFD analysis of void distribution for the BWR fuel bundle test benchmark

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang-Kee; Hwang, Dae-Hyun [Korea Atomic Energy Research Institute (KAERI), 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Jeong, Jae Jun, E-mail: jjjeong@pusan.ac.kr [School of Mechanical Engineering, Pusan National University, Jangjeon-dong, Geumjeong-gu, Busan 609-735 (Korea, Republic of)

    2013-05-15

    Highlights: ► We analyzed subchannel void distributions using subchannel, system and CFD codes. ► The mean error and standard deviation at steady states were compared. ► The deviation of the CFD simulation was greater than those of the others. ► The large deviation of the CFD prediction is due to interface model uncertainties. -- Abstract: The subchannel grade and microscopic void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility have been evaluated with a subchannel analysis code MATRA, a system code MARS and a CFD code CFX-10. Sixteen test series from five different test bundles were selected for the analysis of the steady-state subchannel void distributions. Four test cases for a high burn-up 8 × 8 fuel bundle with a single water rod were simulated using CFX-10 for the microscopic void distribution benchmark. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5 to 25%. The results of the transient calculations were also similar to each other and very reasonable. The CFD simulation reproduced the overall radial void distribution trend which produces less vapor in the central part of the bundle and more vapor in the periphery. However, the predicted variation of the void distribution inside the subchannels is small, while the measured one is large showing a very high concentration in the center of the subchannels. The variations of the void distribution between the center of the subchannels and the subchannel gap are estimated to be about 5–10% for the CFD prediction and more than 20% for the experiment.

  19. Trade, TRIPS, and pharmaceuticals.

    Science.gov (United States)

    Smith, Richard D; Correa, Carlos; Oh, Cecilia

    2009-02-21

    The World Trade Organization's Agreement on Trade-Related Aspects of Intellectual Property Rights (TRIPS) set global minimum standards for the protection of intellectual property, substantially increasing and expanding intellectual-property rights, and generated clear gains for the pharmaceutical industry and the developed world. The question of whether TRIPS generates gains for developing countries, in the form of increased exports, is addressed in this paper through consideration of the importance of pharmaceuticals in health-care trade, outlining the essential requirements, implications, and issues related to TRIPS, and TRIPS-plus, in which increased restrictions are imposed as part of bilateral free-trade agreements. TRIPS has not generated substantial gains for developing countries, but has further increased pharmaceutical trade in developed countries. The unequal trade between developed and developing countries (ie, exporting and importing high-value patented drugs, respectively) raises the issue of access to medicines, which is exacerbated by TRIPS-plus provisions, although many countries have not even enacted provision for TRIPS flexibilities. Therefore this paper focuses on options that are available to the health community for negotiation to their advantage under TRIPS, and within the presence of TRIPS-plus. PMID:19167054

  20. 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.)

  1. 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

  2. Cotton Trip in China

    Institute of Scientific and Technical Information of China (English)

    Wang Ting

    2010-01-01

    @@ During their trip in Beijing,the leadership delegation members,Charles Parker,Harrison Ashley(Vice President of NCC Ginner Services),along with Karin Malmstrom(China Director of CCI)shared a time to accept the interview,giving a general introduction about their China trip and the cotton industry in USA.

  3. 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

  4. 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)

  5. 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

  6. 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.

  7. Aging related degradation in turbine drives and governors for safety related pumps

    International Nuclear Information System (INIS)

    This study is being performed to examine the relationship between time dependent degradation, and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety related pumps. These pumps are located in the auxiliary feedwater (AFW) system for pressurized water reactor (PWR) plants, and the reactor core isolation cooling (RCIC) and high pressure coolant injection (HPCI) systems for boiling water reactor (BWR) facilities. This research has been conducted by examining current information in NPRDS, reviewing licensee event reports, and thoroughly investigating contacts with operating plant personnel, and by personal observation. The reported information was reviewed to determine the cause of the event and the method of discovery. From this data attempts have been made at determining the predictability of events and possible preventive measures that may be implemented. Findings in a recent study on the auxiliary feedwater system (NUREG/CR-5404) indicate that the turbine drive is the single largest contributor to AFW system degradation. Examination of the data, however, show that the turbine itself is a reliable piece of equipment with a good service record. Most of the documented failures are the result of problems with the turbine controls and the mechanical overspeed trip mechanism, which apparently stem from three major causes identified in the paper. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design seem to indicate that this equipment can be a reliable component in safety systems

  8. Lake Mason trip report

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — This document is a summary of the August 31, 1998 trip to Lake Mason to review the hydrology, current conditions, habitat and wildlife. This site visit occurred due...

  9. 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

  10. 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

  11. Safety protection device for BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable to prevent significant reduction in the reactor water level in a case where motor driven feedwater pumps used only at an extremely low frequency are saved and the remaining pumps are failed to be operated. Constitution: Cooling systems upon reactor isolation are extraordinarily operated by using a logic product signal between a trip signal for motor driven feedwater pumps (MDRFP) that are responsible to the small flow rate feed operation and a stop signal for turbine driven feedwater pumps that are responsible to the high flow rate feedwater operation. That is, in a case where motor driven feed water pumps used for a short period of time upon starting should fail to disable the water feeding, the cooling systems upon reactor isolation are started to suppress the reduction in the reactor water level if the condition that the turbine driven feedwater is interrupted is satisfied. (Horiuchi, T.)

  12. 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)

  13. 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

  14. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant. An application of the CSAU methodology using the BNL engineering plant analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.; Cheng, H.S.; Mallen, A.N. [Brookhaven National Lab., Upton, NY (United States); Johnsen, G.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Lellouche, G.S. [Technical Data Services, Chicago, IL (United States)

    1994-03-01

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

  15. 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)

  16. 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

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

    International Nuclear Information System (INIS)

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

  18. Science Activities for School Trips.

    Science.gov (United States)

    Quint, Walter C.

    1980-01-01

    Describes ways in which physics-learning activity packets can be used by high school students participating in trips organized by other departments. Provides an example of physics activities incorporated into an airplane trip to Disney World. (CS)

  19. Guam Commercial Purchases (Trip Ticket)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — DAWR collects Trip Ticket or purchase invoice data from vendors that buy fish directly from the fishermen. Similar to the trip ticket system in Saipan, this is a...

  20. 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

  1. 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.)

  2. 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.)

  3. 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

  4. 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

  5. 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)

  6. 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)

  7. 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

  8. 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)

  9. The return trip effect: Why the return trip often seems to take less time

    OpenAIRE

    van de Ven, N.; van Rijswijk, L.; Roy, M.M.

    2011-01-01

    Three studies confirm the existence of the return trip effect: The return trip often seems shorter than the initial trip, even though the distance traveled and the actual time spent traveling are identical. A pretest shows that people indeed experience a return trip effect regularly, and the effect was found on a bus trip (Study 1), a bicycle trip (Study 2), and when participants watched a video of someone else traveling (Study 3). The return trip effect also existed when another, equidistant...

  10. 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.)

  11. Trip report Rainwater Basin Nebraska

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — This report is a summary a trip to Rainwater Basin Wetland Management District in 1991, and focuses on the hydrology and soil habitat types. It is part of the...

  12. CNMI Commercial Purchases (Trip Ticket)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Commonwealth of Northern Mariana Islands (CNMI), Division of Fish and Wildlife (DFW) collects 'Trip Ticket' or purchase invoice data from vendors that buy fish...

  13. 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)

  14. Heat transfer in rotating serpentine passages with trips skewed to the flow

    Science.gov (United States)

    Johnson, B. V.; Wagner, J. H.; Steuber, G. D.; Yeh, F. C.

    1992-01-01

    Experiments were conducted to determine the effects of buoyancy and Coriolis forces on heat transfer in turbine blade internal coolant passages. The experiments were conducted with a large scale, multi-pass heat transfer model with both radially inward and outward flow. An analysis of the governing flow equations showed that four parameters influence the heat transfer in rotating passages: coolant-to-wall temperature, rotation number, Reynolds number, and radius-to-passage hydraulic diameter ratio. Results were correlated and compared to previous results from similar stationary and rotating models with smooth walls and with trip strips normal to the flow direction. It was concluded that (1) both Coriolis and buoyancy must be considered in turbine blade cooling designs with trip strips, (2) the effects of rotation are markedly different depending upon the flow direction, and (3) the heat transfer with skewed trip strips is less sensitive to buoyancy than the heat transfer models with either smooth or normal trips. Therefore, skewed trip strips rather than normal trip strips are recommended and geometry-specific tests are required for accurate design.

  15. 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

  16. 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.)

  17. Trip Generation Model Based on Destination Attractiveness

    Institute of Scientific and Technical Information of China (English)

    YAO Liya; GUAN Hongzhi; YAN Hai

    2008-01-01

    Traditional trip generation forecasting methods use unified average trip generation rates to determine trip generation volumes in various traffic zones without considering the individual characteristics of each traffic zone.Therefore,the results can have significant errors.To reduce the forecasting error produced by uniform trip generation rates for different traffic zones,the behavior of each traveler was studied instead of the characteristics of the traffic zone.This paper gives a method for calculating the trip efficiency and the effect of traffic zones combined with a destination selection model based on disaggregate theory for trip generation.Beijing data is used with the trip generation method to predict trip volumes.The results show that the disaggregate model in this paper is more accurate than the traditional method.An analysis of the factors influencing traveler behavior and destination selection shows that the attractiveness of the traffic zone strongly affects the trip generation volume.

  18. 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

  19. 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

  20. Evaluation of the root cause for MSR high level trip in Maanshan

    International Nuclear Information System (INIS)

    Reactor trip due to Moisture Separator Reheater (MSR) high water level has been a long time issue for Maanshan nuclear power plant. The operating experience shows that there are five reactor trips due to MSR high water level. Four out of the five reactor trips are generated when Combined Intermediate valve (CIV) no. 1 is closed during CIV closure test. The fifth reactor trip occurs when the reactor power is increasing from 99% to 100%. An extensive root cause analysis has been performed by Taipower Company. It is concluded that the water accumulated in the cross under leg between the exhaust of high pressure turbine and the inlet of MSR was the water source contributing to the MSR high level trip. Although, Maanshan does not have similar trip after the root cause analysis, it is interested to evaluate the proposed root cause from thermal hydraulic point of view. It is also hoped that some useful guidelines can be established. This paper includes a description of the scenario of reactor trips, a summary of the root cause analysis done by Taipower Company, an examination of possible mechanisms, an identification of key parameters and a presentation of major findings. In addition, the applicability of RELAP5/MOD3 under this condition is discussed. (author)

  1. 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).

  2. 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

  3. 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.

  4. BWR - Spent Fuel Transport and Storage with the TNTM9/4 and TNTM24BH Casks

    International Nuclear Information System (INIS)

    The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., the Muehleberg Nuclear Power Plant owner, is involved in this process and has elected to store its BWR spent fuel in a new high capacity dual-purpose cask, the TNeTeM24BH from the COGEMA Logistics/TRANSNUCLEAR TNTM24 family. The Muehleberg BWR spent fuels are transported by road in a medium size shuttle transport cask and then transferred to a heavy transport/storage cask (dry transfer) in the hot cell of ZWILAG site. For that purpose, COGEMA Logistics designed and supplied: - Two shuttle casks, TNTM9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TNTM9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TNTM24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TNTM9/4 round trips are performed, and one TNTM24BH is loaded. 5 additional TNTM24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TNTM24BH high capacity dual purpose cask and the TNTM9/4 transport cask characteristics and capabilities will then be detailed. (authors)

  5. 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)

  6. One-trip drum operating instruction

    Energy Technology Data Exchange (ETDEWEB)

    Ruff, D.T.

    1994-10-01

    The one trip system is a bagless transfer system for egress of waste from gloveboxes into 55 gallon one-trip drums. The contents of this document give an overview of the assembly, loading, and handling of the one-trip drum for use in the WRAP-1 plant.

  7. Tourism, Field Trips and Geographic Education.

    Science.gov (United States)

    Dilsaver, Lary M.

    In university level geography education, both a market and a need exist for "learning trips" combining the elements of a field trip with those of a commercial tour. Planning such trips involves four steps. The first of these is to establish the conditions of the tour, including identification of the topical focus, specific destinations and sites…

  8. One-trip drum operating instruction

    International Nuclear Information System (INIS)

    The one trip system is a bagless transfer system for egress of waste from gloveboxes into 55 gallon one-trip drums. The contents of this document give an overview of the assembly, loading, and handling of the one-trip drum for use in the WRAP-1 plant

  9. Virtual Field Geologic Trip System

    Institute of Scientific and Technical Information of China (English)

    Jian Wang; Linfu Xue; Xiaojun Zhou

    2003-01-01

    Virtual Field Geologic Trip System (VFGTS) constructed by the technique of visualization can efficiently present geologic field information and widely used in the field of geologic education. This paper introduces the developing thinking of VFGTS and discusses the main implement processes. Building VFGTS mainly includes systemically gathering of field geological data, the building of virtual geological world, and displaying of virtual geologic world and human-computer interaction.

  10. 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

  11. 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.)

  12. 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.)

  13. 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.)

  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. Hitachi turbine technology for nuclear applications

    International Nuclear Information System (INIS)

    Hitachi has supplied more than 1200 steam turbines and generators in the past 70 years for both thermal and nuclear applications. Hitachi nuclear steam turbines have been applied to all major reactor types including BWR's and PHWR's (CANDU). Hitachi's recent experience has included supplying the steam turbines for Qinshan Phase III Unit 1 and 2 in China, powered by two CANDU 6 reactors, as well as several ABWR projects in Japan. Hitachi has focused significant R and D efforts on continuous improvement of nuclear steam turbine technology capitalizing on its continuous supply history and sound technical capability. This paper addresses some of the key developments and newest technologies to be employed for new-build nuclear projects, including the ACR-1000 and Enhanced CANDU 6, and focuses on longer Last Stage Blade (LSB) development, Continuous Cover Blades (CCB), and other enhancements in product reliability and performance. (author)

  16. 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)

  17. 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

  18. 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

  19. 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)

  20. 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.)

  1. ECC-D4 Electostatic Oil Cleaner Design for Heavy-Duty Gas Turbine Applications

    OpenAIRE

    Gorur, Murat

    2010-01-01

    The turbine technology improvements from 1980 onwards have considerably increased mechanical and thermal stresses on turbine oils which, cause oil oxidation and thereby turbine oil degradation (Livingstone et al., 2007; Sasaki & Uchiyama, 2002). If the oil degradation problem is ignored, this might result in serious turbine system erratic trips and start-up operational problems (Overgaag et al., 2009). Oil oxidation by-products, in other words, sludge and varnish contaminants, lead stated...

  2. Organizing small-scale scout trip abroad

    OpenAIRE

    Rautala, Camilla

    2016-01-01

    This thesis was commissioned by the Guides and Scouts of Finland (Suomen Partiolaiset – Finlands Scouter ry.) The thesis describes the organising process of a small-scale scout trip to the members of commissioning company. The Guides and Scouts of Finland directs the scout movement in Finland. The small-scale trips to national camps support the organisation’s objectives of educating world citizens. The objective of this thesis project was to plan and implemented well-organised trip to Bal...

  3. Does a satisfying trip result in more future trips with that mode?

    OpenAIRE

    De Vos, Jonas; Schwanen, Tim; Van Acker, Veronique; Witlox, Frank

    2016-01-01

    Previous studies have indicated that travel satisfaction – the experienced emotions during, and cognitive evaluation of, a trip – can be affected by travel mode choice and other trip characteristics. However, as satisfactory trips might improve a person’s attitudes toward the used mode, persons may be more likely to use that same mode for future trips of the same kind. Hence, a cyclical process between travel mode choice and travel satisfaction might occur. In this paper we analyse this proce...

  4. 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

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

    International Nuclear Information System (INIS)

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

  7. RELAP4/MOD6/U4/J3: a JAERI improved version of RELAP4/MOD6 for transient thermal-hydraulic analysis of LWR including effects of BWR core spray

    International Nuclear Information System (INIS)

    The RELAP4/MOD6/U4/J3 code is the latest version of RELAP4/MOD6/Update4 improved in JAERI. The major improvements and modifications included in this version have been carried out aiming at small break LOCA analysis and BWR-LOCA analysis after core spray initiation. For example, a CCFL calculation model and a spray heat transfer model have been added for BWR-LOCA analysis. Using these models, through calculation from the beginning of blowdown to the end of reflood in BWR-LOCA was made practicable. Furthermore, the analyses of operational transients of LWR were facilitated greatly by an addition of a trip reset function. In this report, the description of the improvements and modifications included in this version, the input data description, and the results of two sample problems are contained. (author)

  8. 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.)

  9. Status of the reactor trip in HANARO

    International Nuclear Information System (INIS)

    Unexpected reactor trip in HANARO since the first criticality in February, 1995 was investigated. The total numbers of the reactor trip events were 136 and it was 10.4 cases on average each year. During the early stage of the HANARO operation from 1995 to 1997, unexpected reactor trips were occurred frequently. 67% of the total unexpected reactor trips were occurred in that period, which were 91 cases. That duration was for a power ascension test as one of the reactor performance tests. The unexpected reactor trips were mainly caused by system problems and operators' error. Some cases were caused by electric power failure. The most frequent system problem was originated from fluctuation of the signal of the neutron power measuring system. To prevent the reactor trips by this phenomenon, the circuit of the neutron and thermal power, which was caused when the deviation of the neutron and thermal power is larger than 3MW while the reactor power increases. To reduce reactor the revised one. That has on effect on reducing unexpected reactor trips dramatically. Reactor trips have been occurred by the class-IV power failure or a few problems of system error but never by operators' error since 2000

  10. Status of the reactor TRIP in HANARO

    International Nuclear Information System (INIS)

    Unexpected reactor trip in HANARO since the first criticality in February, 1995 was investigated. The total numbers of the reactor trip events were 136 and it was 10.4 cases on average each year. During the early stage of the HANARO operation from 1995 to 1997, unexpected reactor trips were occurred frequently. 67% of the total unexpected reactor trips were occurred in that period, which were 91 cases. That duration was for a power ascension test as one of the reactor performance tests. The unexpected reactor trips were mainly caused by system problems and operators' error. Some cases were caused by electric power failure. The most frequent system problem was originated from fluctuation of the signal of the neutron power measuring system. To prevent the reactor trips by this phenomenon, the circuit of the reactor protection system was changed. Operators' error were occurred by mismatching of the neutron and thermal power, which was caused when the deviation of the neutron and thermal power is larger than 3 MW while the reactor power increases. To reduce reactor operators' error, operational procedure was revised and the operators have been trained with the revised one. That has on effect on reducing unexpected reactor trips dramatically. Reactor trips have been occurred by the class-IV power failure or a few problems of system error but never by operators' error since 2000. (author)

  11. 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

  12. 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. (

  13. 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

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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

  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. 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.)

  1. 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

  2. Simulation of boiling water reactor one-pump trip transient by SIMULATE-3K

    International Nuclear Information System (INIS)

    To validate models relevant to the transitional in-core 3-D power behavior, the 3-D core kinetics code SIMULATE-3K was applied to an analysis of one-pump trip test performed in a BWR-5 plant which includes large distortion of the power distribution due to the local insertion of control rods. The core boundary conditions required by SIMULATE-3K were calculated by the plant system analysis code RETRAN-3D. It was found that the transitional APRM signal and LPRM signals calculated by SIMULATE-3K agreed well with the measured data. The results showed that the transitional in-core 3-D power behavior can be appropriately predicted by SIMULATE-3K. (author)

  3. Fellows in the Middle: Fabulous Field Trips

    Science.gov (United States)

    West, Mary Lou

    2008-05-01

    Montclair State University's NSF GK-12 Program focuses on grades 7 and 8 in five urban public school districts in northern New Jersey. Each year four fieldtrips are taken by the students, middle school teachers, and graduate student Fellows. Many interdisciplinary hands-on lessons are written for use before, during and after each trip with this year's theme of Earth history. The Sterling Hill Mine trip evoked lessons on geology, economics, crystal structure, density, and pH. A virtual trip (webcam link) to scientists in the rainforest of Panama prompted critical thinking, categorizing layers and animals, and construction of model food webs. In the field trip to the NJ School of Conservation the students will build model aquifers, measure tree heights, and measure stream flow to compare to their Hackensack River. Finally the students will travel to MSU for a Math/Science Day with research talks, lab tours, hands-on activities, and a poster session. In January 2008 seventeen teachers, Fellows, and grant personnel took a field trip to China to set up collaborations with researchers and schools in Beijing and Xi'an, including the Beijing Ancient Observatory. All field trips are fabulous! Next year (IYA) our theme will be planetary science and will feature field trips to the Newark Museum's Dreyfuss Planetarium, BCC Buehler Challenger & Science Center, and star parties. We look forward to invigorating middle school science and mathematics with exciting astronomy. Funded by NSF #0638708

  4. 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)

  5. C. F. Braun. Standard turbine island design, safety analysis report

    International Nuclear Information System (INIS)

    A standard turbine island used with a BWR is described. It consists of the turbine-generator; steam system; condensate storage, cleanup, and transfer systems; control and instrumentation; water treatment plant; make-up demineralizer; potable and waste water systems; and a compressed air system. The turbine-generator is a tandem-compound nuclear-type turbine with one double-flow high-pressure section and a six-flow low-pressure section in three double-flow low-pressure casings. The turbine is direct connected to an 1800 rpm synchronous a-c generator. A combined moisture separator and two-stage reheater is provided. The main steam system delivers the steam generated in a BWR to the main turbine stop valves. The condensate system maintains proper water inventory. Protective features prevent loss of the system due to electrical failure of a component and isolates faults to ensure continuity of a power supply from alternate sources. (U.S.)

  6. Wind turbine

    International Nuclear Information System (INIS)

    The title invention concerns a wind turbine with a rotor, consisting of a number of blades, each with a front edge and an irregular shaped (sawtooth) back edge. This wind turbine aims at reducing the noise pollution of wind turbines. 1 fig

  7. 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.))

  8. 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)

  9. 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)

  10. 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)

  11. 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)

  12. 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.)

  13. 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

  14. 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.)

  15. Steam turbine: Alternative emergency drive for the secure removal of residual heat from the core of light water reactors in ultimate emergency situation

    Energy Technology Data Exchange (ETDEWEB)

    Souza Dos Santos, R. [Instituto de Engenharia Nuclear CNEN/IEN, Cidade Universitaria, Rua Helio de Almeida, 75 - Ilha do Fundiao, 21945-970 Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores / CNPq (Brazil)

    2012-07-01

    In 2011 the nuclear power generation has suffered an extreme probation. That could be the meaning of what happened in Fukushima Nuclear Power Plants. In those plants, an earthquake of 8.9 on the Richter scale was recorded. The quake intensity was above the trip point of shutting down the plants. Since heat still continued to be generated, the procedure to cooling the reactor was started. One hour after the earthquake, a tsunami rocked the Fukushima shore, degrading all cooling system of plants. Since the earthquake time, the plant had lost external electricity, impacting the pumping working, drive by electric engine. When operable, the BWR plants responded the management of steam. However, the lack of electricity had degraded the plant maneuvers. In this paper we have presented a scheme to use the steam as an alternative drive to maintain operable the cooling system of nuclear power plant. This scheme adds more reliability and robustness to the cooling systems. Additionally, we purposed a solution to the cooling in case of lacking water for the condenser system. In our approach, steam driven turbines substitute electric engines in the ultimate emergency cooling system. (authors)

  16. Steam turbine: Alternative emergency drive for the secure removal of residual heat from the core of light water reactors in ultimate emergency situation

    International Nuclear Information System (INIS)

    In 2011 the nuclear power generation has suffered an extreme probation. That could be the meaning of what happened in Fukushima Nuclear Power Plants. In those plants, an earthquake of 8.9 on the Richter scale was recorded. The quake intensity was above the trip point of shutting down the plants. Since heat still continued to be generated, the procedure to cooling the reactor was started. One hour after the earthquake, a tsunami rocked the Fukushima shore, degrading all cooling system of plants. Since the earthquake time, the plant had lost external electricity, impacting the pumping working, drive by electric engine. When operable, the BWR plants responded the management of steam. However, the lack of electricity had degraded the plant maneuvers. In this paper we have presented a scheme to use the steam as an alternative drive to maintain operable the cooling system of nuclear power plant. This scheme adds more reliability and robustness to the cooling systems. Additionally, we purposed a solution to the cooling in case of lacking water for the condenser system. In our approach, steam driven turbines substitute electric engines in the ultimate emergency cooling system. (authors)

  17. Large Pelagic Logbook Trip Survey (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains catch and effort for fishing trips that are taken by vessels with a Federal permit issued for the swordfish and sharks under the Highly...

  18. Abnormal Events for Emergency Trip in HANARO

    International Nuclear Information System (INIS)

    This report gathers abnormal events related to emergency trip of HANARO that happened during its operation over 10 years since the first criticality on February 1995. The collected examples will be utilized to the HANARO's operators as a useful guide

  19. Pre-Trip Notification Database (PTNS)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The PTNS contains pre-trip notification data from vessels participating in the Northeast Multispecies groundfish fishery from 2010 to present and the Longfin squid...

  20. Trip report : Crescent Lake National Wildlife Refuge

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — This trip report is on a visit to Crescent Lake National Wildlife Refuge on September 20 and 21 2001. Wetlands inspected on the Moore Drainage included Martin,...

  1. Lobster Processing and Sales Trip Report Data

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This is a federally mandated log which is required to be mailed in to NMFS after a fishing trip. This data set includes lobster processing and sales information...

  2. TRIPS and Pharmaceuticals: Implications for India

    OpenAIRE

    S. Mehta, Pradeep

    2012-01-01

    On the post-Uruguay Round world trade scenario, after the accords in agriculture and textiles and clothing, the Agreement on Trade-Related Aspects of Intellectual Property Rights (TRIPs) is the issue affecting developing countries like India. One of the seven intellectual properties covered under TRIPs is that of patents. It has been the contentious issue for several reasons. India is committed to amend its patent laws by the year 2005 (for technologies previously unprotected in its market). ...

  3. Trip electrical circuit of the gyrotion

    International Nuclear Information System (INIS)

    The electron cyclotron resonance heating system of INPE/LAP is shown and the trip electrical circuit of the gyrotron is described, together with its fundamental aspects. The trip electrical circuit consists basically of a series regulator circuit which regulates the output voltage level and controls the pulse width time. Besides that, a protection circuit for both tubes, regulator and gyrotron, against faults in the system. (author)

  4. Scrutinized: The TRIPS Agreement and Public Health

    OpenAIRE

    Subhan, Junaid

    2006-01-01

    SUMMARY The World Trade Organization’s (WTO’s) Agreement on Trade-Related Aspects of Intellectual Property Rights (TRIPS) of 1994 seeks to implement a uniform set of intellectual property protection across member nations to provide greater stability in international economic relations. Critics argue that the TRIPS agreement provides unnecessarily strong protection of intellectual property rights which serves to prevent the ill in developing nations from having access to affordable essential m...

  5. Field Trips as Valuable Learning Experiences in Geography Courses

    Science.gov (United States)

    Krakowka, Amy Richmond

    2012-01-01

    Field trips have been acknowledged as valuable learning experiences in geography. This article uses Kolb's (1984) experiential learning model to discuss how students learn and how field trips can help enhance learning. Using Kolb's experiential learning theory as a guide in the design of field trips helps ensure that field trips contribute to…

  6. 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

  7. 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)

  8. Wind Turbines and Coastal Recreation Demand

    OpenAIRE

    Landry, Craig E.; Tom Allen; Todd Cherry; John C. Whitehead

    2010-01-01

    We examine the impact of coastal wind turbines on coastal tourism and recreation for residents of the northern CAMA counties in North Carolina. A combination of telephone and web survey data are used to assess the impact of coastal wind farms on trip behavior and site choice. Most of the respondents to our telephone survey claim to support offshore wind energy development, and independent survey data suggest that the observed levels of support may be indicative of the broader population in th...

  9. 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

  10. 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)

  11. 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

  12. 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)

  13. Steam Turbines

    Science.gov (United States)

    1981-01-01

    Turbonetics Energy, Inc.'s steam turbines are used as power generating systems in the oil and gas, chemical, pharmaceuticals, metals and mining, and pulp and paper industries. The Turbonetics line benefited from use of NASA research data on radial inflow steam turbines and from company contact with personnel of Lewis Research Center, also use of Lewis-developed computer programs to determine performance characteristics of turbines.

  14. 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

  15. Cooperative control scheme for an HVDC system connected to an isolated BWR nuclear power plant

    International Nuclear Information System (INIS)

    This paper describes a cooperative control system to achieve stable operation of an isolated BWR nuclear plant linked to an HVDC system. In the proposed control system, under normal conditions the power plant is controlled according to the generating power reference and the generator frequency deviation is adjusted by converter power control. Such frequency control is also effective in the case of AC-DC system faults. In addition to the frequency control, an overload control is provided with the HVDC system, where the DC transmission power in the sound poles is increased due to a fault detection signal from the faulty pole. Effects of the above mentioned control systems were studied using digital dynamic programs. The sets of simulation results confirmed that in the case of a DC single pole fault, the plant is able to continue operation without any use of the turbine speed control units even for a restarting failure in the faulty pole. In case of a DC two pole fault, the plant is able to continue operation, being assisted by turbine speed control units when restarting in the faulty poles succeeds. In case of an AC three-line to ground fault near the AC terminal of the converter at the sending or receiving end, the system is able to continue stable operation, being supplemented by the turbine control unit when the faulty section of the AC system is isolated by a main or back-up relaying system

  16. Second trip system for NRU research reactor

    International Nuclear Information System (INIS)

    For the past four decades, the NRU research reactor has played an important role at the Chalk River Laboratories, Atomic Energy of Canada Limited, serving as one of its major research and isotope production facilities. To ensure that it continues as an effective facility, compliant with the current safety standards, a comprehensive upgrade program is underway. Adding a second trip system (STS) is part of this upgrade program, aiming at improving the effectiveness and reliability of the overall shutdown function. This document describes the main features and basic principles of the STS.The STS is an independent, seismically qualified trip system, that guarantees reactor shutdown even if the existing trip system fails. It is designed based on 2 out of 3 general coincidence logic, with minimal interferences and changes to the existing system. In addition to the manual trip in the main control room, a remote manual trip is provided in the new Qualified Emergency Response Centre, which is also seismically qualified and always accessible. Thus, for any reason, if the main control room becomes uninhabitable, the reactor still can be manually shut down from this centre. ((orig.))

  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. 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''

  20. 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

  1. 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

  2. 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.

  3. 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)

  4. 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.

  5. 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)

  6. 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)

  7. Turbine system

    Energy Technology Data Exchange (ETDEWEB)

    McMahan, Kevin Weston; Dillard, Daniel Jackson

    2016-05-03

    A turbine system is disclosed. The turbine system includes a transition duct having an inlet, an outlet, and a passage extending between the inlet and the outlet and defining a longitudinal axis, a radial axis, and a tangential axis. The outlet of the transition duct is offset from the inlet along the longitudinal axis and the tangential axis. The turbine system further includes a turbine section connected to the transition duct. The turbine section includes a plurality of shroud blocks at least partially defining a hot gas path, a plurality of buckets at least partially disposed in the hot gas path, and a plurality of nozzles at least partially disposed in the hot gas path. At least one of a shroud block, a bucket, or a nozzle includes means for withstanding high temperatures.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    International Nuclear Information System (INIS)

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

  10. NEWS: A trip to CERN

    Science.gov (United States)

    Ellison, A. D.

    2000-07-01

    the canteen. Over lunch we mixed with physicists of many different nationalities and backgrounds. Figure 1 Figure 1. In the afternoon we visited Microcosm, the CERN visitors centre, and the LEP control room and also the SPS. Here the students learned new applications for much of the physics of standing waves and resonance that they had been taught in the classroom. Later that night, we visited a bowling alley where momentum and collision theory were put into practice. The following morning we returned to CERN and visited the large magnet testing facility. Here again physics was brought to life. We saw superconducting magnets being assembled and tested and the students gained a real appreciation of the problems and principles involved. The afternoon was rounded off by a visit to a science museum in Geneva - well worth a visit, as some of us still use some of the apparatus on display. Friday was our last full day so we visited Chamonix in the northern Alps. In the morning, we ascended the Aiguille de Midi - by cable car. Twenty minutes and 3842 m later we emerged into 50 km h-1 winds and -10 °C temperature, not counting the -10 °C wind chill factor. A crisp packet provided an unusual demonstration of the effects of air pressure (figure 2). Figure 2 Figure 2. The views from the summit were very spectacular though a few people experienced mild altitude sickness. That afternoon the party went to the Mer de Glace. Being inside a 3 million year-old structure moving down a mountain at 3 cm per day was an interesting experience, as was a tot of whisky with 3 million year-old water. Once again the local scenery was very photogenic and the click and whirr of cameras was a constant background noise. Saturday morning saw an early start for the long drive home. Most students - and some staff - took the opportunity to catch up on their sleep. Thanks are due to many people without whom the trip would never have taken place. Anne Craige, Stuart Williams

  11. 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.)

  12. 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

  13. Gas turbines

    Energy Technology Data Exchange (ETDEWEB)

    Farahan, E.; Eudaly, J.P.

    1978-10-01

    This evaluation provides performance and cost data for commercially available simple- and regenerative-cycle gas turbines. Intercooled, reheat, and compound cycles are discussed from theoretical basis only, because actual units are not currently available, except on a special-order basis. Performance characteristics investigated include unit efficiency at full-load and off-design conditions, and at rated capacity. Costs are tabulated for both simple- and regenerative-cycle gas turbines. The output capacity of the gas turbines investigated ranges from 80 to 134,000 hp for simple units and from 12,000 to 50,000 hp for regenerative units.

  14. Round-trip boat on hydrogen

    International Nuclear Information System (INIS)

    The results of a feasibility study on a PEM (polymer-electrolyte membrane) fuel cell (FC) driven electric round-trip boat are presented and discussed. The study concerns the specification of a PEMFC system design, including a list of components. Also technical and environmental aspects are dealt with and compared with traditional battery-driven electric boats and diesel-driven boats

  15. The Educational Value of Field Trips

    Science.gov (United States)

    Greene, Jay P.; Kisida, Brian; Bowen, Daniel H.

    2014-01-01

    The school field trip has a long history in American public education. For decades, students have piled into yellow buses to visit a variety of cultural institutions, including art, natural history, and science museums, as well as theaters, zoos, and historical sites. Schools gladly endured the expense and disruption of providing field trips…

  16. Memorable Experiences of a Science Field Trip.

    Science.gov (United States)

    Knapp, Doug

    2000-01-01

    Attempts to learn more about memorable experiences associated with science field trips by conducting a 1-month and an 18-month evaluation of elementary school students who had participated in an environmental science program at a community park in a Midwestern city. Concludes that students' memories were nonspecific and disassociated from…

  17. No "big trips" for the universe

    CERN Document Server

    Faraoni, V

    2007-01-01

    It has been claimed in several papers that a phantom energy-dominated universe can undergo a ``big trip'', i.e., tunneling through a wormhole that grows faster than the cosmic substratum due to the accretion of phantom energy, and will reappear on the other mouth of the wormhole. We show that such claims are unfounded and contradict the Einstein equations.

  18. WIPP site and vicinity geological field trip

    International Nuclear Information System (INIS)

    The Environmental Evaluation Group (EEG) is conducting an assessment of the radiological health risks to people from the Waste Isolation Pilot Plant (WIPP). As a part of this work, EEG is making an effort to improve the understanding of those geological issues concerning the WIPP site which may affect the radiological consequences of the proposed repository. One of the important geological issues to be resolved is the timing and the nature of the dissolution processes which may have affected the WIPP site. EEG organized a two-day conference of geological scientists, titled Geotechnical Considerations for Radiological Hazard Assessment of WIPP on January 17-18, 1980. During this conference, it was realized that a field trip to the site would further clarify the different views on the geological processes active at the site. The field trip of June 16-18, 1980 was organized for this purpose. This report provides a summary of the field trip activities along with the participants post field trip comments. Important field stops are briefly described, followed by a more detailed discussion of critical geological issues. The report concludes with EEG's summary and recommendations to the US Department of Energy for further information needed to more adequately resolve concerns for the geologic and hydrologic integrity of the site

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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.)

  4. 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)

  5. 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)

  6. 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.)

  7. 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

  8. 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

  9. Southeast Region Headboat Survey-Trip Estimates by Type

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This is a summary of the number of trips by vessel/area/month/trip type through time and is a means of gauging headboat fishery effort and compliance through time.

  10. Automobile trip to Fish Springs, March 16, 1959

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — These memoranda describe trips to Fish Springs National Wildlife Refuge. The purpose of these trips was to access the condition of the land in the Fish Springs area.

  11. 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 (

  12. 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)

  13. Pelton turbines

    CERN Document Server

    Zhang, Zhengji

    2016-01-01

    This book concerns the theoretical foundations of hydromechanics of Pelton turbines from the engineering viewpoint. For reference purposes, all relevant flow processes and hydraulic aspects in a Pelton turbine have been analyzed completely and systematically. The analyses especially include the quantification of all possible losses existing in the Pelton turbine and the indication of most available potential for further enhancing the system efficiency. As a guideline the book therefore supports further developments of Pelton turbines with regard to their hydraulic designs and optimizations. It is thus suitable for the development and design engineers as well as those working in the field of turbo machinery. Many laws described in the book can also be directly used to simplify aspects of computational fluid dynamics (CFD) or to develop new computational methods. The well-executed examples help better understand the related flow mechanics.

  14. Austenite stability in TRIP steels studied by synchrotron radiation

    NARCIS (Netherlands)

    Blondé, R.

    2014-01-01

    TRIP steel is a material providing great mechanical properties. Such steels show a good balance between high-strength and ductility, not only as a result of the fine microstructure, but also because of the well-known TRIP effect. The Transformation Induced-Plasticity (TRIP) phenomenon is the transfo

  15. Wind turbines

    Energy Technology Data Exchange (ETDEWEB)

    Yeoman, J.C. Jr.

    1978-12-01

    This evaluation of wind turbines is part of a series of Technology Evaluations of possible components and subsystems of community energy systems. Wind turbines, ranging in size from 200 W to 10 MW, are discussed as candidates for prime movers in community systems. Estimates of performance characteristics and cost as a function of rated capacity and rated wind speed are presented. Data concerning material requirements, environmental effects, and operating procedures also are given and are represented empirically to aid computer simulation.

  16. Steam turbine

    OpenAIRE

    Vališ, Petr

    2010-01-01

    The master´s thesis concentrates on a project of steam turbine with controlled extraction points destined for a communal waste incineration plant. First, there the history of devices using steam as a moving medium is introduced and than follows the description of computing program, where the calculation was running. The master´s thesis subject consists in the thermodynamic project of turbine vaning, in basic project of gearbox including the check calculations and in technical economic compari...

  17. Using Virtual Field Trips to Connect Students with University Scientists: Core Elements and Evaluation of zipTrips[TM

    Science.gov (United States)

    Adedokun, Omolola A.; Hetzel, Kristin; Parker, Loran Carleton; Loizzo, Jamie; Burgess, Wilella D.; Robinson, J. Paul

    2012-01-01

    Physical field trips to scientists' work places have been shown to enhance student perceptions of science, scientists and science careers. Although virtual field trips (VFTs) have emerged as viable alternatives (or supplements) to traditional physical fieldtrips, little is known about the potential of virtual field trips to provide the same or…

  18. Effects of wind turbines on UHF television reception: field tests in Denmark, November 1991

    International Nuclear Information System (INIS)

    As a result of a planning application for a wind farm comprising 20 wind turbines at Tynewydd Farm, Gilfach Goch in Mid Glamorgan, it became necessary to produce a Report discussing any detrimental effects the proposal might have on UHF television reception. In order to make that Report as definitive as possible, it was decided to carry out field tests on the exact model of wind turbine to be used to Tynewydd. This required a field trip to Denmark, and the opportunity was taken to make measurements on two other models of turbine at the same time. This Report presents the analysis of the results for all three turbines. (Author)

  19. Investigation of transient models and performances for a doubly fed wind turbine under a grid fault

    DEFF Research Database (Denmark)

    Wang, M.; Zhao, B.; Li, H.; Yang, C.; Ye, R.; Chen, Z.

    2011-01-01

    fed induction generator (DFIG), the assessments of the impact on the electrical transient performances were investigated for the doubly fed wind turbine with different representations of wind turbine drive-train dynamics models, different initial operational conditions and different active crowbar...... trip time. Firstly, the different mathematical models of the doubly fed wind turbine were presented, including the electromagnetic transient models of DFIG, a one-mass lumped model, a two-mass shaft flexible model of the wind turbine drive train system, and the power decoupling control strategies of...

  20. Effects of wind turbines on UHF television reception: field tests in Denmark

    International Nuclear Information System (INIS)

    As a result of a planning application for a windfarm comprising 20 wind turbines at Tynewydd Farm, Gilfach Goch in Mid Glamorgan, a report discussing any detrimental effects the proposal might have on u.h.f. television reception was produced. In order to make the report as definitive as possible, it was decided to carry out field tests on the exact model of wind turbine to be used at Tynewydd. This required a field trip to Denmark, and the opportunity was taken to make measurements on two other models of turbine at the same time. This report presents the analysis of the results for all three turbines. (author)

  1. RELAP5/MOD3.3 analysis of reactor trip event in nuclear power plant

    International Nuclear Information System (INIS)

    Measured plant data from various abnormal events or incidents are of great importance for assessing large system thermal-hydraulic computer codes like RELAP5. In the present study the reactor trip, which occurred at Krsko Nuclear Power Plant (NPP) on April 10, 2005, has been analyzed. The purpose of the analysis was to assess the RELAP5/MOD3.3 Patch 03 computer code against plant measured data and validate the RELAP5 input model for Krsko NPP, which is a two-loop Westinghouse pressurized water reactor. The RELAP5 input model delivered by Krsko NPP was used. The event analyzed was a malfunction, which occurred during a power reduction sequence when regular periodic testing of the turbine valves was performed. This caused plant trip. The calculation agrees very well with the plant measured data when operator actions are modelled properly. It was found out that the long term transient evolution is very sensitive to the steam flows from the steam generators after the reactor trip and only proper modelling of these flows gives good quantitative agreement. (author)

  2. Propagation of the trip behavior in the VENUS vertex chamber

    International Nuclear Information System (INIS)

    The high voltage system of the VENUS vertex chamber occasionally trips by a discharge somewhere among cathode electrodes during data taking. This trip behavior induces often additional trips at other electrodes such as the skin and the grid electrodes in the vertex chamber. This propagation mechanism of trips is so complicated in this system related with multi-electrodes. Although the vertex chamber is already installed inside the VENUS detector and consequently the discharge is not able to observe directly, a trial to estimate the propagation has been done using only the information which appears around the trip circuits and the power supply of the vertex chamber. (author)

  3. Operating experience of TAPS 1 and 2 turbine - generators

    International Nuclear Information System (INIS)

    Tarapur Atomic Power Station is a twin unit of BWR type power station with design capacity of 210 MWe each.. The turbine -generators are G.E. USA make model: L4 and are 18 stages single H.P. and four-stage duel L.P. type with tandem compounding and impulse type blades. The Turbine - Generators are axially coupled and units operate at 1500.r.p.m. Generators are 4 pole 12 KV, 248 MV A A/C machines separately excited and are cooled by hydrogen. Though these machines have operated satisfactorily, there were few instances of major components failures, which needed concerted maintenance and surveillance to keep the performance of machines upto the mark. The paper briefly describes about the operating experience of turbine- generators. A few incidents during their operations along with subsequent analyses are also discussed in this paper. (author)

  4. Examination of turbine discs from nuclear power plants

    International Nuclear Information System (INIS)

    Investigations were performed on a cracked turbine disc from the Cooper Nuclear Power Station, and on two failed turbine discs (governor and generator ends) from the Yankee-Rowe Nuclear Power Station. Cooper is a boiling water reactor (BWR) which went into commercial operation in July 1974, and Yankee-Rowe is a pressurized water reactor (PWR) which went into commercial operation in June 1961. Cracks were identified in the bore of the Cooper disc after 41,913 hours of operation, and the disc removed for repair. At Yankee-Rowe two discs failed after 100,000 hours of operation. Samples of the Cooper disc and both Yankee-Rowe disc (one from the governor and one from the generator end of the LP turbine) were sent to Brookhaven National Laboratory (BNL) for failure analysis

  5. 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

  6. Westinghouse inadvertent plant trip reduction program

    International Nuclear Information System (INIS)

    During recent years, the nuclear power industry has become increasingly aware of the high average number of inadvertent plant trips experienced by commercial units in the United States versus similar experience in other countries, most strikingly, Japan. The need to address this problem stems from the potential consequences which include lost generating capacity (and lost revenue) for the utility, the cost of replacement power, reductions in component and plant life expectancies (due to cycling), and customer dissatisfaction. Perhaps another, more subtle, concern is the impact these trips have on the general public and the confidence that it has in the maturity of the commercial nuclear power program in the United States. This paper serves two purposes: first, it summarizes the program Westinghouse has established to reduce the frequency of unplanned automatic scrams, and second, it discusses several observations, made during the implementation of this program at different utilities

  7. Field trips along the Geological Transpyrenean Road

    Science.gov (United States)

    Burgio, Marion; Dekeyser, Dominique; Richert, Jean Paul; Gruneisen, Pierre; Vazquez-Prada, Diego; Lacazedieu, Annie

    2015-04-01

    GeolVal association, www.geolval.fr , in collaboration with Geoambiente association, created a serie of pedagogical explanations about Pyrenees landscapes along the road from France to Spain, linking the towns of Pau, Oloron,Jaca and Mallos de Riglos. It was a ten years work that lasted in 2008. http://www.routetranspyreneenne.com/home_f.php We can now use this "treasure" with tourists, students and mix some landscapes explanations from a road site with a little field trip along mountain in order to "touch" the rocks. We are working about the importance of field in the way the students understand geosciences and try to organize field trips that aim to simplify and make concrete some key concepts. We also began an interactive software version of this work, presented through a pilot application, made in collaboration with Total petroleum firm.

  8. The SMS-GPS-Trip-Method

    DEFF Research Database (Denmark)

    Reinau, Kristian Hegner; Harder, Henrik; Weber, Michael

    2015-01-01

    data, an approach which is not well suited for capturing data on experiences surrounding trips. Currently increasing research is being done on how to incorporate such data in traffic models, and there is therefore a need for a method, which is suited to collect such data. The new method presented in......This article presents a new method for collecting travel behavior data, based on a combination of GPS tracking and SMS technology, coined the SMS–GPS-Trip method. The state-of-the-art method for collecting data for activity based traffic models is a combination of travel diaries and GPS tracking...... this article builds on ideas from experience sampling methods (ESM) and it is well suited specifically for collecting such experience data. Given the use of SMS technology, this method makes it possible to reach a wide range of respondents. The usefulness of the new method is proven on a theoretical...

  9. 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)

  10. Microstructural Development during Welding of TRIP steels

    OpenAIRE

    Amirthalingam, M.

    2010-01-01

    The Advanced High Strength Steels (AHSS) are promising solutions for the production of lighter automobiles which reduce fuel consumption and increase passenger safety by improving crash-worthiness. Transformation Induced Plasticity Steel (TRIP) are part of the advanced high strength steels which offers a high strength and toughness combination with excellent uniform elongation. However, the higher alloying content of these steel limits their weldability and the thermal cycle of a welding proc...

  11. Advanced Tourist Trip Planning Using Hybrid Recommender

    OpenAIRE

    Saso KOCESKI; Petrevska, Biljana

    2013-01-01

    The paper intends to underline that tourists’ and travelers’ behaviour and preferences have changed dramatically specifically in the way how they search for information. In order to create a tailor-made itinerary and enhance the way of planning a tourist trip, the paper suggests introduction of tourism recommendation systems. For this purpose, a methodology is proposed that assists all interested parties in planning their vacation in more intelligent and sophisticated way by generating a pers...

  12. Trip by Colonel and Mrs Hebbert. 1943

    OpenAIRE

    Hebbert, Henry Eric, 1893-1980, Colonel

    2004-01-01

    229 x 186 mm. A hand-drawn map showing the route of the Hebberts' trip through Eritrea to Ethiopia from Khartoum in 1943. In addition to the map, drawn in blue ink, there are annotations in black ink which read: 'Car Mileage Kh-Kh [presumably Khartoum-Khartoum] 3900 miles' and 'Chain dotted generally by car 3900 mile + By Mule (a) Semiam Mts [presumably Simen Mountains] (b) Lalibela'.

  13. Analysis of reactor coolant pump trip

    International Nuclear Information System (INIS)

    The trip of one reactor coolant pump accident is simulated with using the IAEA-developed WWER-1000 simulation computer software (IAEA-Code) in this article. The results are compared with the information presented for Bushehr Nuclear Power Plant for the same scenario (analyzed by the code Dynamika-97). The obtained results are compatible with the Bushehr Nuclear Power Plant's given data and show a good overall agreement between them

  14. TRIPs Agreement, Important Multilateral WTO Treaty

    Directory of Open Access Journals (Sweden)

    Oana-Maria Florescu

    2006-08-01

    Full Text Available This article aims at presenting the content and the frame of the TRIPs. Agreement. It starts by introducing the reader to the terms that defined the world economical climate by the time of the Agreement negociation. Also, it explains the need of having an Agreement on intellectual property rights with impact on the business world. Moreover, the article reviews the main provisions of the Agreement and the most important intellectual property rights.

  15. Your private trips with Carlson Wagonlit Travel

    CERN Multimedia

    Carlson Wagonlit Travel

    2015-01-01

    Your Carlson Wagonlit Travel agency at CERN (building 62) also organizes private trips!     Do not hesitate to contact the “Tourism” team, at your disposal from Monday to Friday from 8:30 a.m. to 4:30 p.m. Phone: 72763. E-mail: cern@carlsonwagonlit.ch. Since 1 January 2015, everyone working at CERN benefits from lower booking fees.

  16. 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

  17. 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

  18. Scrutinized: the TRIPS agreement and public health.

    Science.gov (United States)

    Subhan, Junaid

    2006-07-01

    The World Trade Organization's (WTO's) Agreement on Trade-Related Aspects of Intellectual Property Rights (TRIPS) of 1994 seeks to implement a uniform set of intellectual property protection across member nations to provide greater stability in international economic relations. Critics argue that the TRIPS agreement provides unnecessarily strong protection of intellectual property rights which serves to prevent the ill in developing nations from having access to affordable essential medications. The first recommendation that this paper makes is to provide two sets of intellectual property protection, one that applies to essential medications such as AIDS drugs and certain antibiotics and another that applies to drugs that treat non-life threatening conditions. The second recommendation builds upon the first recommendation: if two sets of intellectual property protection legislation are enacted, patents on essential medications should be restricted to patents on processes rather than the product itself. The third recommendation seeks to amend the language of the TRIPS agreement to make it obligatory for member nations to implement provisions on compulsory licensing within their domestic legislation. PMID:18523624

  19. Wind turbine

    Science.gov (United States)

    Cheney, Jr., Marvin C.

    1982-01-01

    A wind turbine of the type having an airfoil blade (15) mounted on a flexible beam (20) and a pitch governor (55) which selectively, torsionally twists the flexible beam in response to wind turbine speed thereby setting blade pitch, is provided with a limiter (85) which restricts unwanted pitch change at operating speeds due to torsional creep of the flexible beam. The limiter allows twisting of the beam by the governor under excessive wind velocity conditions to orient the blades in stall pitch positions, thereby preventing overspeed operation of the turbine. In the preferred embodiment, the pitch governor comprises a pendulum (65,70) which responds to changing rotor speed by pivotal movement, the limiter comprising a resilient member (90) which engages an end of the pendulum to restrict further movement thereof, and in turn restrict beam creep and unwanted blade pitch misadjustment.

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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

  5. 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.)

  6. 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.)

  7. 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.)

  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. Stainless steels with trip/twip/sbip effect; Nichtrostende Staehle mit TRIP/TWIP/SBIP-Effekt

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A.; Gutte, H.; Jahn, A.; Scheller, P.R. [Institut fuer Eisen- und Stahltechnologie, TU Bergakademie Freiberg (Germany)

    2009-08-15

    Economic austenitic steels with high energy absorption capability are in the focus of worldwide research activities, whereby the steels which show TRIP, TWIP and/or SBIP effects play a crucial role. New austenitic or austenitic-martensitic stainless steels with a high cold workability and energy absorption capability are currently developed and tested in laboratory scale at the Institute of Iron and Steel Technology at the Technical University Bergakademie Freiberg. The mechanical properties of these steels are essentially influenced by the TRIP, TWIP and SBIP effect, becoming evident in hot formed and solution annealed steels as well as in as-cast steels. The TRIP/TWIP/SBIP effects have a significant impact on the toughness and the strength of stainless steels consisting of metastable austenite. The TRIP effect owns a paramount position since it serves for a simultaneous increase of toughness and strength. The influences of alloying elements like manganese or nickel on the TRIP effect are in the centre of the investigations at the Institute of Iron and Steel Technology. These austenitic or austenitc-martensitic stainless steels provide the ability for new applications fields due to their excellent mechanical properties. Exemplary, in the Collaborative Research Centre SFB 799 ''TRIP-Matrix-Composites'', financed through the Deutsche Forschungsgemeinschaft DFG, the suitability of this new class of steels for cast components in ductile and transformation strengthened high performance (metal) ceramic composite materials will be investigated. (Abstract Copyright [2009], Wiley Periodicals, Inc.) [German] Weltweit verstaerken sich die Bemuehungen bei der Suche nach kostenguenstigen austenitischen Stahlgueten mit hohem Energieabsorptionsvermoegen. Dabei zeichnen sich verschiedene Forschungsrichtungen ab, die sich auf Staehle mit TRIP-, TWIP- und SBIP-Effekt{sup 1} oder entsprechenden Kombinationen konzentrieren. Am Institut fuer Eisen- und

  10. A study on the improvement of a Automatic Seismic Trip System in Hanbit Unit no. 1, 2

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hyun Jin; Lee, Jae Yong [Chosun University, Kwangju (Korea, Republic of)

    2014-10-15

    Constructing Automatic Seismic Trip System(ASTS) was the first measure. ASTS is a facility that automatically actuate reactor trip system when earthquake exceeding seismic design criteria of NPP occurs. Installation of ASTS will reinforce emergency response capabilities against earthquake. Currently, all the NPPs in korea equipped with the facility, however Hanbit Unit 2 experienced unexpected transient event during testing the operability of ASTS. This paper include the analysis of transient event, improvement for stable operation of NPP in the case of ASTS actuate and procedure revival for procuring testability of ASTS by reviewing operation experience of Hanbit Unit 2, Though most of the United states and European NPP are not equipped with ASTS facility since low possibility of earthquake occurrence, Korean NPPs faithfully reflected the lessons of Fukushima accident by installing the ASTS facility which leads to more safety and quick emergency response capabilities. This paper comes to conclusion that simultaneous turbine trip signal actuation with reactor trip signal as a countermeasure to prevent repetition of transient event during ASTS functional test by utilizing educational simulator and RETRAN-3D safety analysis.

  11. 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.

  12. 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)

  13. 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)

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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.

  19. Predictors of trips to food destinations

    Directory of Open Access Journals (Sweden)

    Kerr Jacqueline

    2012-05-01

    Full Text Available Abstract Background Food environment studies have focused on ethnic and income disparities in food access. Few studies have investigated distance travelled for food and did not aim to inform the geographic scales at which to study the relationship between food environments and obesity. Further, studies have not considered neighborhood design as a predictor of food purchasing behavior. Methods Atlanta residents (N = 4800 who completed a travel diary and reported purchasing or consuming food at one of five food locations were included in the analyses. A total of 11,995 food-related trips were reported. Using mixed modeling to adjust for clustering of trips by participants and households, person-level variables (e.g. demographics, neighborhood-level urban form measures, created in GIS, and trip characteristics (e.g. time of day, origin and destination were investigated as correlates of distance travelled for food and frequency of grocery store and fast food outlet trips. Results Mean travel distance for food ranged from 4.5 miles for coffee shops to 6.3 miles for superstores. Type of store, urban form, type of tour, day of the week and ethnicity were all significantly related to distance travelled for food. Origin and destination environment, type of tour, day of week, age, gender, income, ethnicity, vehicle access and obesity status were all significantly related to visiting a grocery store. Home neighborhood environment, day of week, type of tour, gender, income, education level, age, and obesity status were all significantly related to likelihood of visiting a fastfood outlet. Conclusions The present study demonstrated that people travel sizeable distances for food and this distance is related to urban. Results suggest that researchers need to employ different methods to characterize food environments than have been used to assess urban form in studies of physical activity. Food is most often purchased while traveling from locations other

  20. Hunton Group core workshop and field trip

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, K.S. [ed.

    1993-12-31

    The Late Ordovician-Silurian-Devonian Hunton Group is a moderately thick sequence of shallow-marine carbonates deposited on the south edge of the North American craton. This rock unit is a major target for petroleum exploration and reservoir development in the southern Midcontinent. The workshop described here was held to display cores, outcrop samples, and other reservoir-characterization studies of the Hunton Group and equivalent strata throughout the region. A field trip was organized to complement the workshop by allowing examination of excellent outcrops of the Hunton Group of the Arbuckle Mountains.

  1. 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.)

  2. Gas turbine installations in nuclear power plants in Sweden

    International Nuclear Information System (INIS)

    At each of the four nuclear power stations in Sweden (Ringhals, Forsmark, Oskarshamn, Barsebaeck) gas turbine generating sets have been installed. These units are normally used for peak load operation dictated of grid and System requirements but they are also connected to supply the electrical auxiliary load of the nuclear plant as reserve power sources. The gas turbines have automatic start capability under certain abnormal conditions (such as reactor trips, low frequency grid etc) but they can also be started manually from several different locations. Starting time is approximately 2- 3 minutes from start up to full load. (author)

  3. Microstructure Evolution during Friction Stir Spot Welding of TRIP steel

    OpenAIRE

    Lomholt, Trine Colding; Somers, Marcel A. J.; da Silva Fanta, Alice Bastos; Pantleon, Karen

    2013-01-01

    TRansformation Induced Plasticity (TRIP) stål er udviklet til anvendelse i bilindustrien på grund af den fremragende kombination af høj styrke og formbarhed. Mikrostrukturen af TRIP stål er en kompleks blanding af mange mikrostruktur-dele: ferrit, bainit, martensit og austenit. TRIP effekten er aktiveret under indflydelse af ekstern mekanisk belastning, der fører til en martensitisk transformation af austenit. Martensit-dannelsen under deformation bidrager til de fremragende mekaniske egenska...

  4. A Label Correcting Algorithm for Dynamic Tourist Trip Planning

    OpenAIRE

    Jin Li; Peihua Fu

    2012-01-01

    One of the most important considerations for tourist in tourism is how to design an optimal trip planning. Selecting the most interesting points of interest and designing a personalized tourist trip is called a tourist trip design problem (TTDP) and, it can be modeled as a orienteering problem. However, most previous researches solve this problem in static network, which leads to unreasonable results. In this paper, to formulate this orienteering problem in a time-dependent network, a mathema...

  5. Chemical behaviour and distribution of volatile radionuclides in a BWR system with forward-pumped heater drains

    International Nuclear Information System (INIS)

    The first BWR plant with forward-pumped feedwater heater drains was studied. Approximately 60% of the iodine activities produced in the coolant (core production plus feedwater return) were transported in the main steam, and about 75% of that amount was condensed in the high pressure turbine and returned to the reactor. The species H2O.I, rather than HOI as previously believed, was probably the major chemical species of iodine activities in steam carryover, the majority of it was condensed at 180 deg C. Similar behavior of the 18F activity as H2O.F was also observed in the steam carryover. All halogen activities in the reactor coolant were found in the anion forms, F-, I-, and I03-. Only about 4% of 13N produced in the coolant was transported to the main steam. Most of the steam-borne 13N was found to be in the cation form in the moisture separator and reheater drains. In the reactor coolant, about 85% of the 13N was found in the anion fraction and about 15% in the cation fraction. An important result of this measurement is that it can be used to predict the transport distribution of 16N radiation in a BWR system under hydrogen water chemistry conditions. (author)

  6. 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)

  7. Trip Oriented Search on Activity Tra jectory

    Institute of Scientific and Technical Information of China (English)

    陈伟; 赵雷; 许佳捷; 刘冠锋; 郑凯; 周晓方

    2015-01-01

    Driven by the flourish of location-based services, trajectory search has received significant attentions in recent years. Different from existing studies that focus on searching trajectories with spatio-temporal information and text de-scriptions, we study a novel problem of searching trajectories with spatial distance, activities, and rating scores. Given a query q with a threshold of distance, a set of activities, a start point S and a destination E, trip oriented search on activity trajectory (TOSAT) returns k trajectories that can cover the activities with the highest rating scores within the threshold of distance. In addition, we extend the query with an order, i.e., order-sensitive trip oriented search on activity trajectory (OTOSAT), which takes both the order of activities in a query q and the order of trajectories into consideration. It is very challenging to answer TOSAT and OTOSAT efficiently due to the structural complexity of trajectory data with rating infor-mation. In order to tackle the problem efficiently, we develop a hybrid index AC-tree to organize trajectories. Moreover, the optimized variant RAC+-tree and novel algorithms are introduced with the goal of achieving higher performance. Extensive experiments based on real trajectory datasets demonstrate that the proposed index structures and algorithms are capable of achieving high efficiency and scalability.

  8. Mediating cultural borders during science field trips

    Science.gov (United States)

    Lebak, Kimberly

    2007-10-01

    In this paper, I analyze teacher and student roles in the teaching and learning of science at an informal learning center, The Outdoor Classroom. As a white middle class informal learning science teacher, I examine my struggles to teach science to students across boundaries of race, class, gender, and experience with the outdoors during field trips. Through the field trip I did not have the time or face-to-face experience to make sense of the students' culture, see their culture in terms of capital, and align my enactment to benefit their learning. Likewise, the students did not have the time or face-to-face experience with me in order to adapt their cultural capital and build the essential stocks of symbolic and social capital. This research demonstrates how the classroom teacher draws upon previous transactions and emotions to successfully engage her students in practices that promote the participation and learning of science. Through creating culturally adaptive ways of transacting, teachers can provide opportunities for their students to generate positive emotional energy and group solidarity in the learning of science at an informal science center.

  9. Ring aggregation pattern of Human Travel Trips

    CERN Document Server

    Wang, Zi-Yang; Zhu, Peng; Qin, Yong; Jia, Li-Min

    2015-01-01

    Although a lot of attentions have been paid to human mobility, the relationship between travel pattern with city structure is still unclear. Here we probe into this relationship by analyzing the metro passenger trip data.There are two unprecedented findings. One, from the average view a linear law exists between the individual's travel distance with his original distance to city center. The mechanism underlying is a travel pattern we called "ring aggregation", i.e., the daily movement of city passengers is just aggregating to a ring with roughly equal distance to city center.Interestingly, for the round trips the daily travel pattern can be regarded as a switching between the home ring at outer area with the office ring at the inner area. Second, this linear law and ring aggregation pattern seems to be an exclusive characteristic of the metro system. It can not be found in short distance transportation modes, such as bicycle and taxi, neither as multiple transportation modes. This means the ring aggregation p...

  10. Hydraulic turbines and auxiliary equipment

    Energy Technology Data Exchange (ETDEWEB)

    Luo Gaorong [Organization of the United Nations, Beijing (China). International Centre of Small Hydroelectric Power Plants

    1995-07-01

    This document presents a general overview on hydraulic turbines and auxiliary equipment, emphasizing the turbine classification, in accordance with the different types of turbines, standard turbine series in China, turbine selection based on the basic data required for the preliminary design, general hill model curves, chart of turbine series and the arrangement of application for hydraulic turbines, hydraulic turbine testing, and speed regulating device.

  11. 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)

  12. 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)

  13. 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

  14. 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)

  15. 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)

  16. 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)

  17. 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)

  18. 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.)

  19. 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.

  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. 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)

  2. Creating Local Field Trips: Seeing Geographical Principles through Empirical Eyes.

    Science.gov (United States)

    Wheeler, James O.

    1985-01-01

    Discusses how instructors can design a local field trip for undergraduate students enrolled in an economic geography class. The purpose of the field trip is to help students observe and interpret familiar scenes in terms of geographical concepts such as central place theory, changing land use, and spatial competition. (RM)

  3. Psychological Sense of Community and Group Cohesion on Wilderness Trips

    Science.gov (United States)

    Breunig, Mary; O'Connell, Tim; Todd, Sharon; Young, Anderson; Anderson, Lynn; Anderson, Dale

    2008-01-01

    A primary purpose of many wilderness trip programs is the development of positive interpersonal relationships and group experiences that lead to enhanced sense of community among group members. Although there is anecdotal evidence to support the development of sense of community on wilderness trips, there is little empirical evidence to support…

  4. Development of Teacher Attitude Scale towards the Field Trip

    Science.gov (United States)

    Tortop, Hasan Said

    2012-01-01

    A field trip is an excursion by group of students with teachers to a place away from classroom such as natural field, science center, and zoo. So, it is an important tool for renewable energy education. This study was carried out to develop a new scale for measuring teacher attitudes towards the field trip. Teacher attitude scale towards the field…

  5. The "Science" Behind a Successful Field Trip to the Zoo

    Science.gov (United States)

    Scott, Catherine Marie; Matthews, Catherine E.

    2011-01-01

    A field trip to the local zoo is often a staple in many elementary school curricula. Many zoos offer free entry to local teachers and their students. Teachers take students on field trips to enrich the curriculum, make connections to what students are learning in school, and provide students with meaningful learning experiences (Kisiel 2007).…

  6. Elementary school children's science learning from school field trips

    Science.gov (United States)

    Glick, Marilyn Petty

    This research examines the impact of classroom anchoring activities on elementary school students' science learning from a school field trip. Although there is prior research demonstrating that students can learn science from school field trips, most of this research is descriptive in nature and does not examine the conditions that enhance or facilitate such learning. The current study draws upon research in psychology and education to create an intervention that is designed to enhance what students learn from school science field trips. The intervention comprises of a set of "anchoring" activities that include: (1) Orientation to context, (2) Discussion to activate prior knowledge and generate questions, (3) Use of field notebooks during the field trip to record observations and answer questions generated prior to field trip, (4) Post-visit discussion of what was learned. The effects of the intervention are examined by comparing two groups of students: an intervention group which receives anchoring classroom activities related to their field trip and an equivalent control group which visits the same field trip site for the same duration but does not receive any anchoring classroom activities. Learning of target concepts in both groups was compared using objective pre and posttests. Additionally, a subset of students in each group were interviewed to obtain more detailed descriptive data on what children learned through their field trip.

  7. The Beginner's Guide to Interactive Virtual Field Trips

    Science.gov (United States)

    Zanetis, Jan

    2010-01-01

    For students, field trips can be the best of both worlds: a welcome and exciting break from day-to-day classroom activities and a memorable, real-world experience that will solidify the curriculum in their minds. Unfortunately, the most desirable trips--those to far-away, enticing destinations--have long been inaccessible to all but a select few,…

  8. Development and formability analysis of TRIP seamless steel tube

    Institute of Scientific and Technical Information of China (English)

    Zhang Zicheng; Zhu Fuxian

    2014-01-01

    In this paper, the production technology of transformation induced plasticity (TRIP) steel was first introduced into the steel tube manufacture field to produce the steel tubes with high strength and plasticity. The TRIP seamless steel tubes with the microstructure of ferrite, bainite, retained austenite and a little martensite were successfully fabricated using a cold-drawn steel tube with two-stage heat treatment technique and continu- ous heat treatment process, respectively. The ring tensile test and cold bend test were carried out to study the formability of the newly developed TRIP seamless steel tube. The results showed that the TRIP seamless steel tubes have a good cold formability, and they are available to be used in the tube hydroforming process. In ad- dition, the equipment of continuous heat treatment developed in the current study can be used to produce TRIP steel tube, and it may serve as an important reference for the industrial production of TRIP steel tube.

  9. Safety aspects of unplanned shutdowns and trips

    International Nuclear Information System (INIS)

    The issue of unplanned shutdowns and trips is receiving increased attention worldwide in view of its importance to plant safety and availability. There exists significant variation in the number of forced shutdowns for nuclear power plants of the same type operating worldwide. The reduction of the frequency of these events will have safety benefits in terms of reducing the frequency of plant transients and the challenges to the safety systems, and the risks of possible incidents. This report provides an insight into the causes of unplanned shutdowns experienced in operating nuclear power plants worldwide, the good practices that have been found effective in minimizing their occurrence, and the measures that have been taken to reduce these events. Specific information on the experiences, approaches and practices of some countries in dealing with this issue is presented in Appendix A

  10. Open Design Architecture for Round Trip Engineering

    Science.gov (United States)

    Beličák, Miroslav; Pokorný, Jaroslav; Richta, Karel

    This chapter introduces a component design of application logic in special type of information system architecture called Open Design Architecture for Round Trip Engineering (ODARTE). This architecture supports model-driven development and integrates information system design with its executable form. It is possible to extract design anytime, modify it, and load back to change the activity and behavior of information system. In this approach, the application logic represents solely functionality and can be described either by sequential model of Windows Workflow Foundation or by UML activity or interaction diagrams. This approach allows creation of flexible and modifiable meta-design of application logic. Finally, an experimental simulation is shown demonstrating the effect of proposal which relates to pilot version of runtime environment for ODARTE support.

  11. User oriented trajectory search for trip recommendation

    KAUST Repository

    Shang, Shuo

    2012-01-01

    Trajectory sharing and searching have received significant attentions in recent years. In this paper, we propose and investigate a novel problem called User Oriented Trajectory Search (UOTS) for trip recommendation. In contrast to conventional trajectory search by locations (spatial domain only), we consider both spatial and textual domains in the new UOTS query. Given a trajectory data set, the query input contains a set of intended places given by the traveler and a set of textual attributes describing the traveler\\'s preference. If a trajectory is connecting/close to the specified query locations, and the textual attributes of the trajectory are similar to the traveler\\'e preference, it will be recommended to the traveler for reference. This type of queries can bring significant benefits to travelers in many popular applications such as trip planning and recommendation. There are two challenges in the UOTS problem, (i) how to constrain the searching range in two domains and (ii) how to schedule multiple query sources effectively. To overcome the challenges and answer the UOTS query efficiently, a novel collaborative searching approach is developed. Conceptually, the UOTS query processing is conducted in the spatial and textual domains alternately. A pair of upper and lower bounds are devised to constrain the searching range in two domains. In the meantime, a heuristic searching strategy based on priority ranking is adopted for scheduling the multiple query sources, which can further reduce the searching range and enhance the query efficiency notably. Furthermore, the devised collaborative searching approach can be extended to situations where the query locations are ordered. The performance of the proposed UOTS query is verified by extensive experiments based on real and synthetic trajectory data in road networks. © 2012 ACM.

  12. Heat transfer in rotating serpentine passages with selected model orientation for smooth or skewed trip walls

    Science.gov (United States)

    Johnson, B. V.; Wagner, J. H.; Steuber, G. D.; Yeh, F. C.

    1993-01-01

    Experiments were conducted to determine the effects of model orientation as well as buoyancy and Coriolis forces on heat transfer in turbine blade internal coolant passages. Turbine blades have internal coolant passage surfaces at the leading and trailing edges of the airfoil with surfaces at angles which are as large as +/- 50 to 60 degrees to the axis of rotation. Most of the previously-presented, multiple-passage, rotating heat transfer experiments have focused on radial passages aligned with the axis of rotation. Results from serpentine passages with orientations 0 and 45 degrees to the axis of rotation which simulate the coolant passages for the mid chord and trailing edge regions of the rotating airfoil are compared. The experiments were conducted with rotation in both directions to simulate serpentine coolant passages with the rearward flow of coolant or with the forward flow of coolant. The experiments were conducted for passages with smooth surfaces and with 45 degree trips adjacent to airfoil surfaces for the radial portion of the serpentine passages. At a typical flow condition, the heat transfer on the leading surfaces for flow outward in the first passage with smooth walls was twice as much for the model at 45 degrees compared to the model at 0 degrees. However, the differences for the other passages and with trips were less. In addition, the effects of buoyancy and Coriolis forces on heat transfer in the rotating passage were decreased with the model at 45 degrees, compared to the results at 0 degrees. The heat transfer in the turn regions and immediately downstream of the turns in the second passage with flow inward and in the third passage with flow outward was also a function of model orientation with differences as large as 40 to 50 percent occurring between the model orientations with forward flow and rearward flow of coolant.

  13. 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

  14. Turbine main engines

    CERN Document Server

    Main, John B; Herbert, C W; Bennett, A J S

    1965-01-01

    Turbine Main Engines deals with the principle of operation of turbine main engines. Topics covered include practical considerations that affect turbine design and efficiency; steam turbine rotors, blades, nozzles, and diaphragms; lubricating oil systems; and gas turbines for use with nuclear reactors. Gas turbines for naval boost propulsion, merchant ship propulsion, and naval main propulsion are also considered. This book is divided into three parts and begins with an overview of the basic mode of operation of the steam turbine engine and how it converts the pressure energy of the ingoing ste

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

    Energy Technology Data Exchange (ETDEWEB)

    Campos Gonzalez, Rina Margarita

    2007-07-15

    This work consists of the integration of three models previously developed which are described widely in Literature: model of the thermo-hydraulic channel, model of the modal neutronic and the model of the recirculation bows. The tool used for this connection of models is the PVM system, Parallel Virtual Machine that allowed paralleling the model by means of the concept of distributed computation. The purpose of making this connection of models is the one of obtaining a more complete tool than better represents the real configuration and the phenomenology of the nucleus of a BWR reactor, thus obtaining better results. In addition to maintaining the flexibility to improve the resulting model at any time, since the very complex or sophisticated models are difficult to improve being impossible to modify the equations they use and can include variables that are not of primary importance in the tackled problem or that mask relations among variables due to the excess of results. Also maintaining the flexibility for adding component of models or systems of the BWR reactor, all of this following the modeling needs. The Swedish Ringhals power plant was chosen to characterize the resulting connected model for counting on a Stability Benchmark that offers the opportunity to count on real plant data. Besides that in case 9 of cycle 14 of this Benchamark oscillations outside phase appeared, which are from great interest because the detection systems that register the average of the power of the nucleus do not detect them. Additionally in this work the model of the recirculation bows as an independent module is obtained in an individual way, since this model belongs to another work and works connected to the reactor vessel. The model of the recirculation bows is able to model several transients of interest, as it is shown in the Appendix A of this work, among which are found the tripping of recirculation pumps or the transference at low or high velocity of them. The scope of the

  16. HIGH EFFICIENCY TURBINE

    OpenAIRE

    VARMA, VIJAYA KRUSHNA

    2012-01-01

    Varma designed ultra modern and high efficiency turbines which can use gas, steam or fuels as feed to produce electricity or mechanical work for wide range of usages and applications in industries or at work sites. Varma turbine engines can be used in all types of vehicles. These turbines can also be used in aircraft, ships, battle tanks, dredgers, mining equipment, earth moving machines etc, Salient features of Varma Turbines. 1. Varma turbines are simple in design, easy to manufac...

  17. Turbine maintenance and modernization

    Energy Technology Data Exchange (ETDEWEB)

    Unga, E. [Teollisuuden Voima Oy, Olkiluoto (Finland)

    1998-12-31

    The disturbance-free operation of the turbine plant plays an important role in reaching good production results. In the turbine maintenance of the Olkiluoto nuclear power plant the lifetime and efficiency of turbine components and the lifetime costs are taken into account in determining the turbine maintenance and modernization/improvement program. The turbine maintenance program and improvement/modernization measures taken in the plant units are described in this presentation. (orig.)

  18. Importance of momentum effects in BWR reactor vessel modeling

    International Nuclear Information System (INIS)

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

  19. 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

  20. 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

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

    International Nuclear Information System (INIS)

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

  2. Development of internal CRD for next generation BWR

    International Nuclear Information System (INIS)

    In order to develop a competitive and high performance Next Generation BWR with fossil power plant, an internal CRD using a heatproof ceramics insulated coil is under development. In case of a 1700MWe next generation BWR, the internal CRDs are installed in a RPV whose size is equivalent to the 1356 MWe ABWR, and there will be no space required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a drywell. This brings further advantages of elimination of RIA (Reactivity Induced Accidents) caused by CR withdrawing under pressure boundary broken, and easy IVR (In Vessel Retention) by vessel bottom cooling in case of a severe accidents. (author)

  3. Development of internal CRD for next generation BWR

    International Nuclear Information System (INIS)

    In order to develop a competitive and high performance Next Generation BWR with fossil power plant, an internal CRD using a heatproof ceramics insulated coil has been developed. In case of a 1700MWe next generation BWR, the internal CRDs are installed in a RPV whose size is equivalent to the 1356 MWe ABWR, and there will be no space required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a drywell. This brings further advantages of elimination of RIA (Reactivity Induced Accidents) caused by CR withdrawing under pressure boundary broken, and easy IVR (In Vessel Retention) by vessel bottom cooling in case of severe accidents. (authors)

  4. Study on fast spectrum BWR core for actinide recycle

    International Nuclear Information System (INIS)

    A study has been performed on neutronic and thermal characteristics of a fast spectrum BWR core with tight fuel lattice for an innovative fuel cycle system named BARS (BWR with an Advanced Recycle System) aiming at Pu multi-recycling and MA's (Minor Actinides) burning. The coolant void reactivity tends to shift to positive direction due to the fast neutron spectrum, so the neutron streaming channel and axial enrichment distribution was introduced in the core design to overcome the tendency. It was found that negative void reactivity is attained for whole burnup cycle by using Monte Carlo code. As for the thermal characteristics, the critical power measurement test was performed in order to get the database and the result was compared with the Arai's critical power correlation. As the result, Arai's correlation can predict our critical power test data considerably. Through our core design study and benchmark tests, we will make the conceptual design of BARS core successfully. (authors)

  5. Neutron noise analysis of BWR using time series analysis

    International Nuclear Information System (INIS)

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

  6. Corrosion of Inconel X750 in simulated BWR core environment

    International Nuclear Information System (INIS)

    The corrosion tests of Inconel X750 in the simulated core environment by using an irradiated test loop were performed. From the experiments, the followings were clarified; 1. corrosion rate was proportional to minus cubic root of the time and the value was evaluated to be 19±4 mdm after one year, 2. the value was markedly high compared to other reported experimental data, but it had a good coincidence with the value evaluated from activity balance data in actual BWR plants, 3. the most important chemical specie which affected the corrosion behavior was identified to be hydrogen peroxide, and 4. the rate of specimens pre-filmed in the high temperature atmosphere at 700deg C for 5 hrs was reduced to be one third or less than that of specimens pre-oxidized in steam at 390deg C for 13 hrs which was the same condition for actual BWR fuel springs. (author)

  7. REVIEW AND PROSPECT OF HIGH STRENGTH LOW ALLOY TRIP STEEL

    Institute of Scientific and Technical Information of China (English)

    L. Li; P. Wollants; Y.L. He; B.C. De Cooman; X.C. Wei; Z.Y. Xu

    2003-01-01

    Research status of high strength low alloy TRIP (transformation induced plasticity)steels for automobile structural parts is briefly described. Composition and microstructure factors especially the morphology, size and volume fraction of retained austenite,which largely influence the strength and ductility of the steel, are reviewed and discussed one after another. Modelling of the inter-critical annealing and martempering processes as well as the designing of the TRIP steel aided by commercial software are introduced. Some special aspects of the dynamic mechanical properties of TRIP steel are firstly reported.

  8. Electrochemical potential measurements under simulated BWR water chemistry conditions

    International Nuclear Information System (INIS)

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

  9. 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

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

    International Nuclear Information System (INIS)

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

  11. 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

  12. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

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

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

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

  14. Peer Pressure Road Trip 2012 - A Report

    Science.gov (United States)

    Popnoe, Olivia

    2012-10-01

    Each year since 2005 the Angelo State University SPS ``Peer Pressure Team'' has travelled for a week in May to promote physics and other sciences for grades K-12. This year, the Peer Pressure Team visited schools in West Texas and New Mexico. This was the first year the team has traveled outside of Texas, inspired by the SPS 2012 theme ``Physics Beyond Borders.'' The purpose of the Road Tours is to encourage an interest in science in children in grades K-12 and give experience in presenting and explaining physics to the undergraduate volunteers. The schools attended are carefully selected based on socioeconomic need and an underrepresentation in physics, as well as other sciences. During this week, the Peer Pressure Team saw about 1,500 students over 10 shows. Surveys were given to teachers post-demonstration to have students complete and return so the Team could gauge the effectiveness of the outreach program. In addition, the surveys allow the Team to determine which demonstrations had the most impact. We present the details of this year's road trip with the results and analysis of the survey data.

  15. You Can't Reach for the Stars if You are Tripping Over the Ground! (Preventing Slips, Trips, and Falls)

    Science.gov (United States)

    Miller, Darcy; Raysich, Mark; Kirkland, Mary

    2016-01-01

    Although there are very few mishaps related to ground, vehicle or payload processing at the Kennedy Space Center (KSC), employees have experienced a significant number of injuries due to slips, trips, and falls outside of performing flight processing operations. Slips, trips, and falls are major causes of occupational injuries at KSC, the National Aeronautics and Space Administration (NASA), and in general industry. To help KSC employees avoid these injuries, and allow them to be fully productive, KSC launched an initiative in 2013 to reduce slips, trips, and falls. This initiative is based on a four-part model focusing on DATA analysis, HAZARD awareness, PREVENTIVE methods, and BALANCE.

  16. 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

  17. [August 1992 trip report : Kirwin NWR and Rainwater Basin WMD

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — This trip report evaluates the current years farming and grazing results for Kirwin National Wildlife Refuge and Rainwater Basin Wetland Management District.

  18. Analisis Sympathetic Trip pada Penyulang Ungasan-Bali Resort, Bali

    Directory of Open Access Journals (Sweden)

    Cakasana Alif Bathamantri

    2012-09-01

    Full Text Available Sistem pengaman dalam tenaga listrik dimaksudkan untuk melokalisir gangguan agar tidak meluas sesuai dengan cakupan daerah pengaman. Salah satu kegagalan pengaman melokalisir gangguan disebut sympathetic trip yaitu kegagalan rele pada penyulang 20 kV di gardu induk dimana penyulang yang tidak terganggu, akan ikut trip dengan penyulang yang terganggu. Dalam tugas akhir ini akan dibahas mengenai analisa penyebab terjadinya peristiwa sympathetic trip pada penyulang Ungasan-Bali Resort 20kV di Bali karena di tempat inilah tercatat sering terjadi gangguan sympathetic trip. Besarnya arus kapasitif, setting ground fault relay dan koordinasi rele pengaman semua akan di analisa pada tugas akhir ini. Untuk analisa koordinasi rele, perhitungan dilakukan menggunakan program ETAP 7.0.

  19. Probabilistic determination of neutronic trip setpoints at Cernavoda NPP, Romania

    International Nuclear Information System (INIS)

    This paper describes the methodology used for probabilistic determination of the Regional Overpower tripsetpoints at Cernavoda Nuclear Power Plant Unit 1 (a CANDU project). The shutdown systems use in-core flux detectors to protect against local power increases that can lead to fuel dryout and melting. These flux detectors are spread throughout the reactor core and are assigned to 3 independent trip channels per shutdown systems. The design criteria is that each shutdown system must trip the reactor with a confidence of 98% on a 2 of 2 logic, i.e. assuming the best trip channel unavailable. Parameter considered in the calculation of the trip-setpoints (TSP) are affected by random and systematic error. The probabilistic approach described here was developed by AECL specialists and allows error margins of ∼ 10%, lower than the deterministic margin of 15-20%.(author)

  20. Hawaii Longline Fishery Trip Expenditure (2004 to present)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This is a time-series dataset of trip expenditure data for the Hawaii-based longline fleet for the period August 2004 to present. The data collection includes 10...

  1. Customer satisfaction with individual shopping trip experiences in grocery retailing

    DEFF Research Database (Denmark)

    Esbjerg, Lars; Grunert, Klaus G; Jensen, Birger Boutrup;

    individual shopping trips is a prerequisite for developing customer loyalty. However, there is surprisingly little research focusing on satisfaction with individual shopping trips. On the contrary, satisfaction is normally conceptualised and studied as an overall evaluation of a given retailer based on all...... mechanisms may be at work than in other retail settings such as themed flagship stores, which are visited less regularly and are oriented more towards creating hedonic shopping experiences. This paper develops a conceptual framework for analysing customer satisfaction with individual shopping trip...... experiences in grocery retailing. The framework makes at least two important contributions to the literature. First, it focuses on customer satisfaction with individual shopping trips whereas previous research and theoretical frameworks have addressed either overall satisfaction with the retailer, service...

  2. [July 1999 trip report : Kirwin National Wildlife Refuge

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — This report summarizes a trip to Kirwin National Wildlife Refuge in 1999. It provides general background on the refuge, and reviews recreational activities, habitat...

  3. My trip to Britain,my greatest inspiration

    Institute of Scientific and Technical Information of China (English)

    陈嘉瑜

    2015-01-01

    During a summer vacation I went to Britain to visit our sister school there with the teachers and students delegation in our school for a cross-cultural communication trip and we stayed there for about a month.What I saw and experienced during the trip broadened my horizon and had a great effect on me and it’s really a wonderful and unforgettable experience in my life.

  4. Systematic comparison of trip distribution laws and models

    OpenAIRE

    Lenormand, Maxime; Bassolas, Aleix; Ramasco, José J.

    2015-01-01

    Trip distribution laws are basic for the travel demand characterization needed in transport and urban planning. Several approaches have been considered in the last years. One of them is the so-called gravity law, in which the number of trips is assumed to be related to the population at origin and destination and to decrease with the distance. The mathematical expression of this law resembles Newton's law of gravity, which explains its name. Another popular approach is inspired by the theory ...

  5. Modelling leisure day trips between Berlin and its surrounding

    OpenAIRE

    Beuck, Ulrike

    2005-01-01

    The paper describes the development, application and validation of a quantitative spatial model for leisure day trips from Berlin into the surrounding regions. Leisure activities can contribute substantially to the economies of rural regions like the surroundings of Berlin, but these activities are also often very sensitive to environmental quality, which is usually negatively affected by the traffic caused by the leisure travelers...A model of the spatial distribution of leisure day trips ca...

  6. Systematic comparison of trip distribution laws and models

    OpenAIRE

    Lenormand, Maxime; Bassolas, Aleix; José J. Ramasco

    2016-01-01

    Trip distribution laws are basic for the travel demand characterization needed in transport and urban planning. Several approaches have been considered in the last years. One of them is the so-called gravity law, in which the number of trips is assumed to be related to the population at origin and destination and to decrease with the distance. The mathematical expression of this law resembles Newton's law of gravity, which explains its name. Another popular approach is inspired by the theory ...

  7. Accommodation of the spinal cat to a tripping perturbation

    Directory of Open Access Journals (Sweden)

    ReggieEdgerton

    2012-05-01

    Full Text Available Adult cats with a complete spinal cord transection at T12-T13 can relearn over a period of days-to-weeks how to generate full weight-bearing stepping on a treadmill or standing ability if trained specifically for that task. In the present study, we assessed short-term (msec-min adaptations by repetitively imposing a mechanical perturbation on the hindlimb of chronic spinal cats by placing a rod in the path of the leg during the swing phase to trigger a tripping response. The kinematics and EMG were recorded during control (10 steps, trip (1 to 60 steps with various patterns and then release (without any tripping stimulus, 10 to 20 steps sequences. Our data show that the activation patterns and kinematics of the hindlimb in the step cycle immediately following the initial trip (mechanosensory stimulation of the dorsal surface of the paw was modified in a way that increased the probability of avoiding the obstacle in the subsequent step. This indicates that the spinal sensorimotor circuitry reprogrammed the trajectory of the swing following a perturbation prior to the initiation of the swing phase of the subsequent step, in effect “attempting” to avoid the re-occurrence of the perturbation. The average height of the release steps was elevated compared to control regardless of the pattern and the length of the trip sequences. In addition, the average impact force on the tripping rod tended to be lower with repeated exposure to the tripping stimulus. EMG recordings suggest that the semitendinosus, a primary knee flexor, was a major contributor to the adaptive tripping response. These results demonstrate that the lumbosacral locomotor circuitry can modulate the activation patterns of the hindlimb motor pools within the time frame of single step in a manner that tends to minimize repeated perturbations. Furthermore, these adaptations remained evident for a number of steps after removal of the mechanosensory stimulation.

  8. A Field Trip without Buses: Connecting Your Students to Scientists through a Virtual Visit

    Science.gov (United States)

    Adedokun, Omolola; Parker, Loran Carleton; Loizzo, Jamie; Burgess, Wilella; Robinson, J. Paul

    2011-01-01

    One school offers zipTrips as an alternative to actual field trips taken outside of school. ZipTrips are web- and broadcast-delivered electronic field trips that include online videos, lesson plans, and a live, 45-minute interactive program consisting of four core components: an in-studio audience, live interaction with scientists, prerecorded…

  9. Field Trips Put Chemistry in Context for Non-Science Majors

    Science.gov (United States)

    Peterman, Keith E.

    2008-01-01

    Field trips can provide excellent real-world learning situations for students in non-science major chemistry courses. The field trips described in this article are accessible, most trips can be completed within a scheduled three-hour laboratory time period, and they can be conducted at minimal cost. These field trips significantly enhanced student…

  10. Language Travel or Language Tourism: Have Educational Trips Changed So Much?

    Science.gov (United States)

    Laborda, Jesus Garcia

    2007-01-01

    This article points out the changes in organization, students and language learning that language trips, as contrasted with educational trips (of which language trips are a subgroup) have gone through in the last years. The article emphasizes the need to differentiate between language trips and language tourism based on issues of additional…

  11. Development and formability analysis of TRIP seamless steel tube

    Institute of Scientific and Technical Information of China (English)

    Zhang Zicheng; Zhu Fuxian

    2014-01-01

    In this paper,the production technology of transformation induced plasticity (TRIP) steel was first introduced into the steel tube manufacture field to produce the steel tubes with high strength and plasticity. The TRIP seamless steel tubes with the microstructure of ferrite,bainite,retained austenite and a little martensite were successfully fabricated using a cold-drawn steel tube with two-stage heat treatment technique and continu-ous heat treatment process,respectively. The ring tensile test and cold bend test were carried out to study the formability of the newly developed TRIP seamless steel tube. The results showed that the TRIP seamless steel tubes have a good cold formability,and they are available to be used in the tube hydroforming process. In ad-dition,the equipment of continuous heat treatment developed in the current study can be used to produce TRIP steel tube,and it may serve as an important reference for the industrial production of TRIP steel tube.

  12. Round trip landing missions to the asteroid Eros, 1981 opportunity

    Science.gov (United States)

    Zambo, G. A.

    1971-01-01

    A comprehensive, Del V (magnitude of impulsive velocity-change vector) study of short staytime (0 to 30 days), short total trip time (10 to 160 days) round trip landing missions (fast missions) is presented. The characteristics of the following selected round trip landing missions are described: mission A (fast mission) with 120-day total trip time, 0-day stay-time, and 73,000-fps total mission Del V; mission B (opposition class) with 380-day total trip time, 0-day staytime, and 52,000-fps total mission Del V; and mission C (conjuction class) with 540-day total trip time, 175-day staytime, and 39,000-fps total mission Del V. A study of large-tank Agena, chemical propulsion stage (CPS), and nuclear stage was conducted. A payload of 5000 pounds was assumed. The propulsion requirements for mission A were found to be two large-tank Agenas and ten CPSs (or alternatively, two large-tank Agenas and five nuclear stages) while mission B required two large-tank Agenas and two CPSs. Mission C can be performed with one large-tank Agena and one CPS.

  13. Analysis of the Unexpected HANARO Trip History(1996 - 2005)

    International Nuclear Information System (INIS)

    Since its first criticality in February 1995, the HANARO has been utilized for nuclear fuel and material irradiation tests, radioisotope production, neutron beam applications, neutron activation analyses, and a neutron transmutation doping, etc. In the early HANARO operations, the unexpected reactor trip occurred frequently during a power ascension test which was one of the reactor performance tests. They were most due to a system problem and an operator's error. Some cases originated from an experimenter error and the failure of an electric power supply. To reduce the unexpected reactor trips due to system problems, the neutron power measuring system was improved. A retraining of the operators was performed to reduce the human errors. When the reactor power increases, the reactor is tripped in the case that the difference of the neutron and thermal power is larger than 3 MW. To prevent a reactor trip by the thermal and neutron power mismatch, the neutron power is calibrated during an increasing power when the neutron power and thermal power differences are above 1.5MW. In this paper, the cases of unexpected reactor trips are analyzed from 1996 to 2005 and the efforts to reduce the unexpected reactor trips and their effects are described

  14. Hybrid intelligent monironing systems for thermal power plant trips

    Science.gov (United States)

    Barsoum, Nader; Ismail, Firas Basim

    2012-11-01

    Steam boiler is one of the main equipment in thermal power plants. If the steam boiler trips it may lead to entire shutdown of the plant, which is economically burdensome. Early boiler trips monitoring is crucial to maintain normal and safe operational conditions. In the present work two artificial intelligent monitoring systems specialized in boiler trips have been proposed and coded within the MATLAB environment. The training and validation of the two systems has been performed using real operational data captured from the plant control system of selected power plant. An integrated plant data preparation framework for seven boiler trips with related operational variables has been proposed for IMSs data analysis. The first IMS represents the use of pure Artificial Neural Network system for boiler trip detection. All seven boiler trips under consideration have been detected by IMSs before or at the same time of the plant control system. The second IMS represents the use of Genetic Algorithms and Artificial Neural Networks as a hybrid intelligent system. A slightly lower root mean square error was observed in the second system which reveals that the hybrid intelligent system performed better than the pure neural network system. Also, the optimal selection of the most influencing variables performed successfully by the hybrid intelligent system.

  15. Small wind turbines

    CERN Document Server

    Wood, David

    2011-01-01

    Small Wind Turbines provides a thorough grounding in analysing, designing, building, and installing a small wind turbine. Small turbines are introduced by emphasising their differences from large ones and nearly all the analysis and design examples refer to small turbines.The accompanying software includes MATLAB(R) programs for power production and starting performance, as well as programs for detailed multi-objective optimisation of blade design. A spreadsheet is also given to help readers apply the simple load model of the IEC standard for small wind turbine safety. Small Wind Turbines repr

  16. WIPP site and vicinity geological field trip

    International Nuclear Information System (INIS)

    The Environmental Evaluation Group is conducting an assessment of the radiological health risks to people from the Waste Isolation Pilot Plant. As a part of this work, EEG is making an effort to improve the understanding of those geological issues concerning the WIPP site which may affect the radiological consequences of the proposed repository. One of the important geological issues to be resolved is the timing and the nature of the dissolution processes which may have affected the WIPP site. EEG organized a two-day conference of geological scientists, on January 17-18, 1980. On the basis of the January conference and the June field trip, EEG has formed the following conclusions: (1) it has not been clearly established that the site or the surrounding area has been attacked by deep dissolution to render it unsuitable for the nuclear waste pilot repository; (2) the existence of an isolated breccia pipe at the site unaccompanied by a deep dissolution wedge, is a very remote possibility; (3) more specific information about the origin and the nature of the brine reservoirs is needed. An important question that should be resolved is whether each encounter with artesian brine represents a separate pocket or whether these occurrences are interconnected; (4) Anderson has postulated a major tectonic fault or a fracture system at the Basin margin along the San Simon Swale; (5) the area in the northern part of the WIPP site, identified from geophysical and bore hole data as the disturbed zone, should be further investigated to cleary understand the nature and significance of this structural anomaly; and (6) a major drawback encountered during the discussions of geological issues related to the WIPP site is the absence of published material that brings together all the known information related to a particular issue

  17. User Oriented Trajectory Search for Trip Recommendation

    KAUST Repository

    Ding, Ruogu

    2012-09-08

    Trajectory sharing and searching have received significant attention in recent years. In this thesis, we propose and investigate the methods to find and recommend the best trajectory to the traveler, and mainly focus on a novel technique named User Oriented Trajectory Search (UOTS) query processing. In contrast to conventional trajectory search by locations (spatial domain only), we consider both spatial and textual domains in the new UOTS query. Given a trajectory data set, the query input contains a set of intended places given by the traveler and a set of textual attributes describing the traveler’s preference. If a trajectory is connecting/close to the specified query locations, and the textual attributes of the trajectory are similar to the traveler’s preference, it will be recommended to the traveler. This type of queries can enable many popular applications such as trip planning and recommendation. There are two challenges in UOTS query processing, (i) how to constrain the searching range in two domains and (ii) how to schedule multiple query sources effectively. To overcome the challenges and answer the UOTS query efficiently, a novel collaborative searching approach is developed. Conceptually, the UOTS query processing is conducted in the spatial and textual domains alternately. A pair of upper and lower bounds are devised to constrain the searching range in two domains. In the meantime, a heuristic searching strategy based on priority ranking is adopted for scheduling the multiple query sources, which can further reduce the searching range and enhance the query efficiency notably. Furthermore, the devised collaborative searching approach can be extended to situations where the query locations are or- dered. Extensive experiments are conducted on both real and synthetic trajectory data in road networks. Our approach is verified to be effective in reducing both CPU time and disk I/O time.

  18. User Oriented Trajectory Search for Trip Recommendation

    KAUST Repository

    Ding, Ruogu

    2012-07-08

    Trajectory sharing and searching have received significant attention in recent years. In this thesis, we propose and investigate the methods to find and recommend the best trajectory to the traveler, and mainly focus on a novel technique named User Oriented Trajectory Search (UOTS) query processing. In contrast to conventional trajectory search by locations (spatial domain only), we consider both spatial and textual domains in the new UOTS query. Given a trajectory data set, the query input contains a set of intended places given by the traveler and a set of textual attributes describing the traveler’s preference. If a trajectory is connecting/close to the specified query locations, and the textual attributes of the trajectory are similar to the traveler’s preference, it will be recommended to the traveler. This type of queries can enable many popular applications such as trip planning and recommendation. There are two challenges in UOTS query processing, (i) how to constrain the searching range in two domains and (ii) how to schedule multiple query sources effectively. To overcome the challenges and answer the UOTS query efficiently, a novel collaborative searching approach is developed. Conceptually, the UOTS query processing is conducted in the spatial and textual domains alternately. A pair of upper and lower bounds are devised to constrain the searching range in two domains. In the meantime, a heuristic searching strategy based on priority ranking is adopted for scheduling the multiple query sources, which can further reduce the searching range and enhance the query efficiency notably. Furthermore, the devised collaborative searching approach can be extended to situations where the query locations are ordered. Extensive experiments are conducted on both real and synthetic trajectory data in road networks. Our approach is verified to be effective in reducing both CPU time and disk I/O time.

  19. Analysis of a BWR MKI utilizing advanced BMI-2104 computational techniques to calculate source terms

    International Nuclear Information System (INIS)

    The effects of incorporating code modifications to more realistically represent accident phenomenology were examined with specific focus on the BWR MKI AE γ-sequence. Results indicate that the source terms are only a fraction of those reported in BMI-2104 and that core/concrete interaction is not a significant source term contributor for the BWR MKI containments. (author)

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

    International Nuclear Information System (INIS)

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

  1. Aerodynamics of wind turbines

    DEFF Research Database (Denmark)

    Hansen, Martin Otto Laver

    Aerodynamics of Wind Turbines is the established essential text for the fundamental solutions to efficient wind turbine design. Now in its third edition, it has been substantially updated with respect to structural dynamics and control. The new control chapter now includes details on how to design...... Wind Turbines (VAWT). Topics covered include increasing mass flow through the turbine, performance at low and high wind speeds, assessment of the extreme conditions under which the turbine will perform and the theory for calculating the lifetime of the turbine. The classical Blade Element Momentum...... method is also covered, as are eigenmodes and the dynamic behaviour of a turbine. The book describes the effects of the dynamics and how this can be modelled in an aeroelastic code, which is widely used in the design and verification of modern wind turbines. Furthermore, it examines how to calculate the...

  2. Aerodynamics of Wind Turbines

    DEFF Research Database (Denmark)

    Hansen, Martin Otto Laver

    Aerodynamics of Wind Turbines is the established essential text for the fundamental solutions to efficient wind turbine design. Now in its second edition, it has been entirely updated and substantially extended to reflect advances in technology, research into rotor aerodynamics and the structural...... response of the wind turbine structure. Topics covered include increasing mass flow through the turbine, performance at low and high wind speeds, assessment of the extreme conditions under which the turbine will perform and the theory for calculating the lifetime of the turbine. The classical Blade Element...... Momentum method is also covered, as are eigenmodes and the dynamic behavior of a turbine. The new material includes a description of the effects of the dynamics and how this can be modeled in an aeroelastic code, which is widely used in the design and verification of modern wind turbines. Further, the...

  3. 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.

  4. Gas Turbine Performance Deterioration

    OpenAIRE

    Verlo, Simen Berg

    2010-01-01

    IToday, gas turbines are generally used to produce electricity and drive natural gas compressors on offshore installations. The performance of gas turbines is reduced gradually, and this lead to reduced production capacity of the platform. Degradation of performance of a gas turbine also carry with them higher fuel consumption and higher emissions. Degradation rate of gas turbines have proven to be very site-specific, but compressor fouling is generally the largest contributor to reduced per...

  5. Backpressure Steam turbine

    OpenAIRE

    Chrástek, Pavel

    2012-01-01

    The diploma works demonstrates the thermodynamic calculation and design of the geometry flow profile of the back-pressure steam turbine for specified inlet and outlet conditions of steam. Next target of the work is to design the drawing of the turbine axial cross section. Structuring of this diploma work calculation is following: - regulating stage thermodynamic calculation - turbine stages calculation - equalizing-balancing piston calculation - entire turbine intrinsic efficiency and power r...

  6. Turbulence and wind turbines

    DEFF Research Database (Denmark)

    Brand, Arno J.; Peinke, Joachim; Mann, Jakob

    2011-01-01

    The nature of turbulent flow towards, near and behind a wind turbine, the effect of turbulence on the electricity production and the mechanical loading of individual and clustered wind turbines, and some future issues are discussed.......The nature of turbulent flow towards, near and behind a wind turbine, the effect of turbulence on the electricity production and the mechanical loading of individual and clustered wind turbines, and some future issues are discussed....

  7. Wind Turbine Blade Design

    OpenAIRE

    Richard J. Crossley; Peter J. Schubel

    2012-01-01

    A detailed review of the current state-of-art for wind turbine blade design is presented, including theoretical maximum efficiency, propulsion, practical efficiency, HAWT blade design, and blade loads. The review provides a complete picture of wind turbine blade design and shows the dominance of modern turbines almost exclusive use of horizontal axis rotors. The aerodynamic design principles for a modern wind turbine blade are detailed, including blade plan shape/quantity, aerofoil selection ...

  8. Field Trips and Their Effect on Student Achievement and Attitudes: A Comparison of Physical versus Virtual Field Trips to the Indian River Lagoon

    Science.gov (United States)

    Garner, Lesley C.; Gallo, Michael A.

    2005-01-01

    This study examined the effect of physical and virtual field trips on undergraduate, nonscience majors. No significant differences were seen in achievement, attitudes, learning styles, interactions between field trip and learning styles, or students' ability to answer questions at different levels. Results imply that both field trips promote…

  9. FIX-II. Loca-blowdown heat transfer and pump trip experiments. Summary report of phase 1: Design of experiments

    International Nuclear Information System (INIS)

    FIX-II is a loss of coolant blowdown heat transfer experiment, performed under contract for The Swedish Nuclear Power Inspectorate, SKI. The purpose of the experiments is to provide measurements from simulations of a pipe rupture on an external recirculation line in a Swedish BWR. Pump trips in BWRs with internal recirculation pumps will also be simulated. The existing FIX-loop at the Thermal Engineering Laboratory of Studsvik Energiteknik AB will be modified and used for the experiments. Components are included to simulate the steam dome, downcomer, two recirculation lines with one pump each, lower plenum, core (36-rod full length bundle), control rod guide tubes, core bypass, upper plenum and steam separators. The results of the first phase of the project are reported here. The following tasks are included in Phase 1: reactor reference analysis, scaling calculations of the FIX loop, development of fuel rod simulators, design of test section and test loop layout and proposal for test program. Further details of the work and results obtained for the different sub-projects are published in a number ofdetailed reports. (author)

  10. 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.

  11. 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.

  12. 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

  13. 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)

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

    OpenAIRE

    Javier Ortiz-Villafuerte; Andrés Rodríguez-Hernández; Enrique Araiza-Martínez; Luis Fuentes-Márquez; Jorge Viais-Juárez

    2013-01-01

    RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liqui...

  15. Characterization studies of BWR-4 neutron noise analysis spectra

    International Nuclear Information System (INIS)

    Neutron noise analysis measurements were made in three BWR-4 reactors under full-power conditions to determine the noise characterization spectra of the reactors with two different instrument-tube cooling configurations. Both configurations were designed to prevent flow-induced vibration of the instrument tubes and subsequent damage of fuel channel boxes caused by impacts of the tubes with the boxes. Noise spectra from these three reactors were compared with spectra previously obtained prior to changing the instrument-tube cooling configuration, and no evidence of impacting was found

  16. In-situ testing of BWR closure head studs

    International Nuclear Information System (INIS)

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

  17. 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)

  18. 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)

  19. 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

  20. Void effects on BWR Doppler and void reactivity feedback

    International Nuclear Information System (INIS)

    The significance of steam voids and control rods on the Doppler feedback in a gadolinia shimmed BWR is demonstrated. The importance of bypass voids when determining void feedback is also shown. Calculations were done using a point model, i.e., feedback was expressed in terms of reactivity coefficients which were determined for individual four-bundle configurations and then appropriately combined to yield reactor results. For overpower transients the inclusion of the void effect of control rods is to reduce Doppler feedback. For overpressurization transients the inclusion of the effect of bypass void wil increase the reactivity due to void collapse. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-02-01

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

  2. Development of underwater surface inspection system for BWR suppression chamber

    International Nuclear Information System (INIS)

    In order to inspect underwater paints surface of BWR suppression chamber as an in-service inspection, the underwater surface inspection system has been developed, which consists of magnetic crawler, crud recovery nozzles and special camera unit. The system can move on wall surface of pressure suppression chamber with magnetic crawler to inspect surface paints films under the muddy water with special camera unit after removing surface cruds. Based on inspection results, the paint films inspection criteria to assess degradation of paints films were proposed. (T. Tanaka)

  3. The return trip is felt shorter only postdictively: A psychophysiological study of the return trip effect [corrected].

    Directory of Open Access Journals (Sweden)

    Ryosuke Ozawa

    Full Text Available The return trip often seems shorter than the outward trip even when the distance and actual time are identical. To date, studies on the return trip effect have failed to confirm its existence in a situation that is ecologically valid in terms of environment and duration. In addition, physiological influences as part of fundamental timing mechanisms in daily activities have not been investigated in the time perception literature. The present study compared round-trip and non-round-trip conditions in an ecological situation. Time estimation in real time and postdictive estimation were used to clarify the situations where the return trip effect occurs. Autonomic nervous system activity was evaluated from the electrocardiogram using the Lorenz plot to demonstrate the relationship between time perception and physiological indices. The results suggest that the return trip effect is caused only postdictively. Electrocardiographic analysis revealed that the two experimental conditions induced different responses in the autonomic nervous system, particularly in sympathetic nervous function, and that parasympathetic function correlated with postdictive timing. To account for the main findings, the discrepancy between the two time estimates is discussed in the light of timing strategies, i.e., prospective and retrospective timing, which reflect different emphasis on attention and memory processes. Also each timing method, i.e., the verbal estimation, production or comparative judgment, has different characteristics such as the quantification of duration in time units or knowledge of the target duration, which may be responsible for the discrepancy. The relationship between postdictive time estimation and the parasympathetic nervous system is also discussed.

  4. Appraisal of boundary layer trips for landing gear testing

    Science.gov (United States)

    McCarthy, Philip; Feltham, Graham; Ekmekci, Alis

    2013-11-01

    Dynamic similarity during scaled model testing is difficult to maintain. Forced boundary layer transition via a surface protuberance is a common method used to address this issue, however few guidelines exist for the effective tripping of complex geometries, such as aircraft landing gears. To address this shortcoming, preliminary wind tunnel tests were performed at Re = 500,000. Surface transition visualisation and pressure measurements show that zigzag type trips of a given size and location are effective at promoting transition, thus preventing the formation of laminar separation bubbles and increasing the effective Reynolds number from the critical regime to the supercritical regime. Extension of these experiments to include three additional tripping methods (wires, roughness strips, CADCUT dots) in a range of sizes, at Reynolds number of 200,000 and below, have been performed in a recirculating water channel. Analysis of surface pressure measurements and time resolved PIV for each trip device, size and location has established a set of recommendations for successful use of tripping for future, low Reynolds number landing gear testing.

  5. Understanding intra-urban trip patterns from taxi trajectory data

    Science.gov (United States)

    Liu, Yu; Kang, Chaogui; Gao, Song; Xiao, Yu; Tian, Yuan

    2012-10-01

    Intra-urban human mobility is investigated by means of taxi trajectory data that are collected in Shanghai, China, where taxis play an important role in urban transportation. From the taxi trajectories, approximately 1.5 million trips of anonymous customers are extracted on seven consecutive days. The globally spatio-temporal patterns of trips exhibit a significant daily regularity. Since each trip can be viewed as a displacement in the random walk model, the distributions of the distance and direction of the extracted trips are investigated in this research. The direction distribution shows an NEE-SWW-dominant direction, and the distance distribution can be well fitted by an exponentially truncated power law, with the scaling exponent β = 1.2 ± 0.15. The observed patterns are attributed to the geographical heterogeneity of the study area, which makes the spatial distribution of trajectory stops to be non-uniform. We thus construct a model that integrates both the geographical heterogeneity and distance decay effect, to interpret the observed patterns. Our Monte Carlo simulation results closely match to the observed patterns and thus validate the proposed model. According to the proposed model, in a single-core urban area, the geographical heterogeneity and distance decay effect improve each other when influencing human mobility patterns. Geographical heterogeneity leads to a faster observed decay, and the distance decay effect makes the spatial distribution of trips more concentrated.

  6. Evidence, explanations, and recommendations for teachers' field trip strategies

    Science.gov (United States)

    Rebar, Bryan

    Field trips are well recognized by researchers as an educational approach with the potential to complement and enhance classroom science teaching by exposing students to unique activities, resources, and content in informal settings. The following investigation addresses teachers' field trip practices in three related manuscripts: (1) A study examining the details of teachers' pedagogical strategies intended to facilitate connections between students' experiences and the school curricula while visiting an aquarium; (2) A study documenting and describing sources of knowledge that teachers draw from when leading field trips to an aquarium; (3) A position paper that reviews and summarizes research on effective pedagogical strategies for field trips. Together these three pieces address key questions regarding teachers' practices on field trips: (1) What strategies are teachers employing (and not employing) during self-guided field trips to facilitate learning tied to the class curriculum? (2) What sources of knowledge do teachers utilize when leading field trips? (3) How can teachers be better prepared to lead trips that promote learning? The Oregon Coast Aquarium served as the field trip site for teachers included in this study. The setting suited these questions because the aquarium serves tens of thousands of students on field trips each year but provides no targeted programming for these students as they explore the exhibits. In other words, the teachers who lead field trips assume much of the responsibility for facilitating students' experience. In order to describe and characterize teachers' strategies to link students' experiences to the curriculum, a number of teachers (26) were observed as they led their students' visit to the public spaces of the aquarium. Artifacts, such as worksheets, used during the visit were collected for analysis as well. Subsequently, all teachers were surveyed regarding their use of the field trip and their sources of knowledge for

  7. Steam drum pressure controller response for AHWR turbine trip operational transient

    International Nuclear Information System (INIS)

    Proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. Passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level with no primary coolant pumps. This paper deals with details of mathematical modeling and predicts different thermal-hydraulics parameters obtained during the transient study. The analysis has been performed using a two-fluid code RELAP5/MOD3.2. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, qualities and flow in different part of Primary Heat Transport (PHT) system. On the basis of the results, the requirements and parameters of steam drum pressure control systems will be finalized. (author)

  8. Turbine Imaging Technology Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Moursund, Russell A.; Carlson, Thomas J.

    2004-12-31

    The goal of this project was to identify and evaluate imaging alternatives for observing the behavior of juvenile fish within an operating Kaplan turbine unit with a focus on methods to quantify fish injury mechanisms inside an operating turbine unit. Imaging methods are particularly needed to observe the approach and interaction of fish with turbine structural elements. This evaluation documents both the opportunities and constraints for observing juvenile fish at specific locations during turbine passage. The information may be used to acquire the scientific knowledge to make structural improvements and create opportunities for industry to modify turbines and improve fish passage conditions.

  9. Wind Turbine Blade Design

    Directory of Open Access Journals (Sweden)

    Richard J. Crossley

    2012-09-01

    Full Text Available A detailed review of the current state-of-art for wind turbine blade design is presented, including theoretical maximum efficiency, propulsion, practical efficiency, HAWT blade design, and blade loads. The review provides a complete picture of wind turbine blade design and shows the dominance of modern turbines almost exclusive use of horizontal axis rotors. The aerodynamic design principles for a modern wind turbine blade are detailed, including blade plan shape/quantity, aerofoil selection and optimal attack angles. A detailed review of design loads on wind turbine blades is offered, describing aerodynamic, gravitational, centrifugal, gyroscopic and operational conditions.

  10. TRIBES - A CPC (Core Protection Calculator)/CEAC (Control Element Assembly Calculator) TRIp Buffer Expert System

    International Nuclear Information System (INIS)

    This paper discusses the development of a small IBM personal computer based expert system to diagnose reactor trips initiated by the Core Protection Calculator, CPC, and the Control Element Assembly Calculator, CEAC. The expert system interprets information from trip buffers created by the CPC and CEAC to determine the reason for the reactor trip. A trip buffer consists of a number of numeric point IDs and their values at the time of the CPC or CEAC channel trip. An expert system approach was chosen because trip buffer interpretation requires a CPC expert to determine the cause of a trip, the knowledge is amenable to formulation in rules and the justification subsystem of an expert system allows a novice user to follow the program's logic to the trip conclusion. The TRIp Buffer Expert System, TRIBES, inference engine is written in TURBO-PROLOG

  11. 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)

  12. Microstructure characterization of Friction Stir Spot Welded TRIP steel

    DEFF Research Database (Denmark)

    Lomholt, Trine Colding; Adachi, Yoshitaka; Peterson, Jeremy;

    2012-01-01

    Transformation Induced Plasticity (TRIP) steels have not yet been successfully joined by any welding technique. It is desirable to search for a suitable welding technique that opens up for full usability of TRIP steels. In this study, the potential of joining TRIP steel with Friction Stir Spot...... Welding (FSSW) is investigated. The aim of the study is to investigate whether acceptable welds can be produced, and additionally, to obtain an understanding of the microstructural changes during welding. The microstructure was investigated with a combination of microscopical techniques with the aim of...... identifying the transformations occurring during welding. Reflected light microscopy, scanning electron microscopy, and electron backscatter diffraction were among the methods applied for detailed investigations. The microstructure adjacent to the welds can generally be subdivided in two thermo...

  13. Mechanical and Microstructural Characterization of an Aluminum Bearing Trip Steel

    Science.gov (United States)

    Monsalve, Alberto; Guzmán, Alexis; De Barbieri, Flavio; Artigas, Alfredo; Carvajal, Linton; Bustos, Oscar; Garza-Montes-de Oca, Nelson F.; Colás, Rafael

    2016-03-01

    The mechanical properties and microstructural characteristics of a steel able to sustain the TRIP-effect were studied. The material was prepared by taking in mind the partial substitution of silicon by aluminum following a processing route that included hot forging, hot and cold rolling, intercritical annealing, and a final bainitic isothermal treatment. The mechanical properties that were obtained resulted to be above those of commercial a 780 TRIP steel. The TRIP phenomenon was confirmed by the change in retained austenite before and after deforming the steel; X-ray diffraction was used to evaluate the volume content of retained austenite. Formability of the steel under study can be rationalized in terms of the texture developed in the material.

  14. Microstructure and Mechanical Properties of High Manganese TRIP Steel

    Institute of Scientific and Technical Information of China (English)

    ZHAO Jin-long; XI Yan; SHI Wen; LI Lin

    2012-01-01

    Microstructure evolution and mechanical properties of newly designed 0.1C-6Mn-0.5Si-1Al TRIP-aided steels under different annealing conditions and the effects of matrix microstructure before intercritical annealing on the final microstructure were studied by means of X-ray diffraction(XRD),scanning electron microcopy(SEM),dilatometric simulation,optical microstructure(OM) and tensile testing in this work.The experimental results indicate that the TRIP steel with Mn of 6% could form a considerable amount of retained austenite with good TRIP effect after a simple intercritical annealing treatment,and the matrix microstructure before intercritical annealing treatment can greatly affect the final microstructure.The original microstructure of the ferritic matrix steel was eliminated,while annealed martensite was remained from the martensite matrix steel under the same intercritical annealing conditions

  15. Mechanical and Microstructural Characterization of an Aluminum Bearing Trip Steel

    Science.gov (United States)

    Monsalve, Alberto; Guzmán, Alexis; De Barbieri, Flavio; Artigas, Alfredo; Carvajal, Linton; Bustos, Oscar; Garza-Montes-de Oca, Nelson F.; Colás, Rafael

    2016-06-01

    The mechanical properties and microstructural characteristics of a steel able to sustain the TRIP-effect were studied. The material was prepared by taking in mind the partial substitution of silicon by aluminum following a processing route that included hot forging, hot and cold rolling, intercritical annealing, and a final bainitic isothermal treatment. The mechanical properties that were obtained resulted to be above those of commercial a 780 TRIP steel. The TRIP phenomenon was confirmed by the change in retained austenite before and after deforming the steel; X-ray diffraction was used to evaluate the volume content of retained austenite. Formability of the steel under study can be rationalized in terms of the texture developed in the material.

  16. Are short daily trips compensated by higher leisure mobility?

    DEFF Research Database (Denmark)

    Næss, Petter

    2006-01-01

    Studies in several cities have shown that inner-city residents travel shorter distances and use cars less for local transport than suburbanites do. However, according to some authors, a low daily amount of travel is likely to be compensated through more extensive leisure mobility at weekends and on...... holidays. On the basis of a study of residential location and travel in the Copenhagen metropolitan area, this paper addresses the phenomenon of compensatory travel. For travel within ‘weekend trip distance’ from the residence, inner-city living appears to have a certain compensatory effect in the form of...... a higher frequency of medium-distance leisure trips. Probably, this reflects a shortage of nature in the immediate surroundings of the dwelling as well as less leisure time tied to gardening and house maintenance. These compensatory trips imply a slight reduction of the transport-reducing effect of...

  17. Transforming an Exposure trip to Botanical Expedition: Introducing Ecological Research thru Exposure Trip in an Eco-tourism Site

    Directory of Open Access Journals (Sweden)

    Bernardo C. Lunar

    2014-10-01

    Full Text Available – Fieldtrips can be considered as one of the three avenues through which science can be taught - through formal classroom teaching, practical work and field trips. An exposure trip at Bangkong Kahoy Valley Field Study Center was arranged for a class of BS Biology and BS Education students enrolled in Ecology Course. This approach purposefully transformed the usual exposure trip from being a casual site visit into a focused and productive learning experience. This transformation from exposure trip to a botanical expedition has exceeded the initial activity goals. Rather than a day off from learning, the time spent at the study center has been a meaningful opportunity to engage students in an active ecological research project while delivering valuable science content. Employing the descriptive survey design, the learning gains of the students were assessed and students were directed to do a guided reflection writing using the ORID Model of Focused Conversation. The learning gains and reflections of the students confirmed that students can collaboratively develop focused research questions, make meaning from a variety of sources, carry out a vegetation analysis and conduct surveys on socio-economic status, plant resource utilization and ecotourism assessment of the host community. As students prepared for their trip and synthesized their learning afterward, they were able to come up with very impressive and scientifically sound research outputs.

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

    International Nuclear Information System (INIS)

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 1018 n/cm2) in the TRIGA Mark III Salazar reactor and separately with Ni+3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A2). (Author)

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  1. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. 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.

  4. 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)

  5. 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)

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

    International Nuclear Information System (INIS)

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

  7. Analysis of NEACRP 3D BWR core transient benchmark

    International Nuclear Information System (INIS)

    NEACRP BWR cold water injection benchmark is analyzed by two codes: TRAC-BF1/SKETCH-N code system by JAERI, Japan and TRAB-3D code by VTT Energy, Finland. Basic features of the codes are described. Neutronics modules of the codes apply nodal methods; separate calculations are performed to compare their accuracy. Thermal-hydraulics modules are significantly different: TRAC-BF1 uses two-phase two-fluid model, while TRAB-3D applies drift-flux model with four separated equations. A representative set of the global and local reactor parameters is given for both the steady-state and transient conditions. TRAB-3D calculations have been performed with two slip correlations: EPRI and the simple Zuber-Findley correlation. A comparison of the two TRAB results shows the importance of the slip model on some computed reactor parameters. The results of the TRAC-BF1/SKETCH-N and TRAB-3D codes are in a close agreement, especially when the advanced EPRI correlation is used in the TRAB-3D code. The presented data can be useful for assessment of other BWR codes. (author)

  8. BWR Mark II ex-vessel corium interaction analyses

    International Nuclear Information System (INIS)

    This report describes the results of a series of studies conducted to investigate the behavior of core debris within a BWR Mark II containment. These studies focused on the interaction of core debris with concrete and steel structures (downcomers and inpedestal floor drains) within the drywell, the transport of debris through these drains and downcomers into the wetwell, and on debris-water reactions within the wetwell. Estimates of the conditions under which debris would penetrate the in-pedestal drain lines, the time-dependent behavior of the debris within the drain lines, and the amount of debris which might enter the suppression pool via these drain lines are provided. An assessment of the conditions under which the upper lip of the downcomers would be expected to fail (i.e. melt) due to exposure to hot core debris is presented. Finally, the unique characteristics of debris water interactions in Mark II containments are discussed, the existing knowledge base regarding core-concrete debris-water interactions is summarized, and an evaluation of the applicability of the MELCOR 1.80 code's debris-water interaction model to BWR Mark II's is presented

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  11. Efforts for optimization of BWR core internals replacement

    International Nuclear Information System (INIS)

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

  12. 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)

  13. 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

  14. Aging assessment of BWR control rod drive systems

    Energy Technology Data Exchange (ETDEWEB)

    Greene, R.H.

    1991-01-01

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

  15. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

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

  16. 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

  17. The development of cause analysis system for CPCS trip using the rule-base deduction

    International Nuclear Information System (INIS)

    The Core Protection Calculator System(CPCS) was developed to initiate a Reactor Trip under the circumstance of certain transients by Combustion Engineering Company. The major function of the CPCS is to generate contact outputs for the Departure from Nucleate Boiling Ratio(DNBR) Trip and Local Power Density(LPD) Trip. But in CPCS the trip causes can not be identified, only trip status is displayed. It may take much time and efforts for plant operator to analyse the trip causes of CPCS. So, the Cause Analysis System for CPCS(CASCPCS) has been developed using the rule-base deduction method to aid the operators in Nuclear Power Plant

  18. Evaluation of pressure transitories in BWR type reactors using the BWRDYN code; Evaluacion de transitorios de presion en reactores tipo BWR usando el codigo BWRDYN

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez P, J.A. [ESIME, Unidad Profesional Azcapotzalco, Av. de las Granjas 682, 02550 Mexico D.F. (Mexico)]. e-mail: jrodriguez@ipn.mx

    2007-07-01

    Several simulations of pressure transitory for a nucleo electric power station with BWR/4 type reactor were carried out. The simulated pressure transitories were made for the Peach Bottom 2 Nucleo electric central. Also, it was carried out for the same Plant the simulation of the turbine shot with derivation to the main condenser, of the reference case (benchmark) outlined by the Organization for the Cooperation and the Economic Development and of the Commission Regulatory in Nuclear matter of the United States of America. As tool to carry out the simulations of the transitory ones, the BWRDYN code developed by the Japan Energy Research Institute was used. Among the main suppositions and models that it includes the BWRDYN code its can be mentioned: a) that of punctual kinetics that calculates the neutron flow; for the calculation of the fuel temperature, this it is divided in nodes in the radial and axial directions, the wrapper is considered like a region in the radial direction; c) the pressure is supposed that it is uniform inside the reactor vessel; and d) the thermal hydraulic pattern of the reactor vessel is divided in five regions and the core is divided in several nodes to take into account the distribution of holes in the axial direction. The modeling of the control systems of the feeding water system is also included, of the pressure regulator and of the recirculation system. The systems of what is known as plant balance are also modeled. The numeric results of the simulations provide valuable information of the behavior of the nucleo electric central. The obtained results of the simulation of the reference case agree acceptably with the measurements data, when comparing them with the measurements made in the Peach Bottom 2 Central. The obtained results of each simulation are fundamental to evaluate the transitory one, as well as to delineate the sequence and the impact of diverse events that they happen during the same one transitory. In the case of the

  19. 汽轮机停机故障分析%Breakdown Analysis of Turbine

    Institute of Scientific and Technical Information of China (English)

    刘康宁

    2015-01-01

    When the Over Speed Control (OPC) acts very often, the pressure switches of safety fuel will be released, which maybe lead to the turbine tripped. Some steam leak out from the governor valve flange face. The work temperature of LVDT is very high. The coils of LVDT are sometimes short out. The actuator isout of control inducing the turbine tripped. The improvement suggestions are proposed in this paper.%某电厂超速限制OPC动作时,安全油压力开关复位导致停机。调节阀漏汽,LVDT工作温度过高,线圈短路,油动机失控导致停机。本文提出了设计改进建议。

  20. Support information for slips, trips and falls from height offshore

    International Nuclear Information System (INIS)

    This document contains support information for the Slips, Trips and Falls from Height Offshore Final Report produced by BOMEL for the Health and Safety Executive's Offshore Division (BOMEL Reference C919/05/171R Rev B). The project was requested in response to a concern that slips, trips and falls from height (STFS) consistently account for a significant proportion of all accidents offshore. The work has provided a detailed understanding of the factors which influence the likelihood of STFs and presents a strategy to reduce such accidents offshore. An important part of the strategy is guidance to enable a practical assessment of STF risk. (author)

  1. What drives people? Analyzing leisure-shopping trip decisions making

    OpenAIRE

    DE CEUNYNCK, Tim; KUSUMASTUTI, Diana; HANNES, Els; JANSSENS, Davy; Wets, Geert

    2011-01-01

    Because of the strong increase in the number of leisure-shopping trips, a shift towards more sustainable leisure-shopping behaviour is desirable. This can be attained by having a better insight into people’s reasoning in choosing a transport mode and shopping location for this type of activities. Thus, this paper highlights individuals’ leisure-trip decision-making processes. The uniqueness of this study is the use of a large sample group, consisting of 221 respondents. A Computer-Based Causa...

  2. Bake hardening behavior of TRIP and DP steels

    Institute of Scientific and Technical Information of China (English)

    Jicheng Zhang; Renyu Fu; Mei Zhang; Rendong Liu; Xicheng Wei; Lin Li

    2008-01-01

    The bake hardening (BH) behavior of transformation-induced plasticity (TRIP) and dual-phase (DP) steels after different prestrains was studied. The experimental results indicate that TRIP steel exhibits good BH ability while DP steel does not, and prestrain displays a strong effect on the BH values of both steels. The comparison of microstructures of the two steels showed that the hard second phase in the matrix might be harmful to the BH ability. For deformed specimens, baking resulted in a loss of uniform elongation, but there was no obvious decrease in uniform elongation for unprestrained specimens.

  3. Controlling the feedwater flow in a BWR. Examples from Forsmark 2; Regleringen av matarvattenfloedet i en BWR. Med exempel fraan Forsmark 2

    Energy Technology Data Exchange (ETDEWEB)

    Bergdahl, Bengt-Goeran; Oguma, Ritsuo (GSE Power Systems AB, Nykoeping (Sweden))

    2009-03-15

    An investigation of the feedwater controller at Forsmark 2 has been performed. The investigation is based on signal analysis of measurement signals recorded during operation of the plant during different tests. The feedwater controller consists of the water level controller, the flow controller and the condenser balance controller. The overall goal of the feedwater control is to maintain constant water level (level controller) in the reactor and at the same time balance the water levels in the two condensers (condenser balance controller) to avoid that one condenser is full of water while the other one is operated with too low level. There is also a feed forward of the difference between steam flow and feedwater flow (flow controller) for each turbine system with the aim to reduce the fluctuation in reactor water level. The relation in strength between the three controllers is such that the level controller is the strongest followed by the condenser balance controller and finally the flow controller. Tests with trip of the feedwater pump and automatic start of the spare pump in each turbine system indicates a fast reduction in reactor water level that is restored after the transient in the control system. The transient in water level is stable without oscillations. However, it takes about 100 s before the reactor water level is restored. The function of the flow controller has been questioned by the authors. It does not take the action that is expected when a disturbance takes place in the difference between steam and feedwater flow. In addition to this principal weakness there is an offset in the feedwater controller output for feedwater flow 22 that reduces the contribution in flow control that is expected during the introduction of a disturbance. This offset should be adjusted during instrument maintenance of the feedwater controller. The PIP parameters for the level controller are gain factors and time constants. These have been evaluated with the aid of

  4. Reactor and turbine building layout of the high performance light water reactor

    International Nuclear Information System (INIS)

    Based on the information generated within the European funded project ''High Per-formance Light Water Reactor Phase 2'', a general plant layout has been developed. The central building is the reactor building, in which the containment and safety sys-tems are located. The reactor building is with app. 90.000 m3 considerably smaller compared to other BWR buildings, thus providing a huge potential for cost savings. The turbine building with app 250,000 m3 is of approximately the same size like for existing BWRs. (orig.)

  5. Corrosion fatigue in LP steam turbine blading - experiences, causes and appropriate measures; Korrosionsutmattning i aangturbinskovlar - Erfarenheter, inverkande faktorer och moejliga aatgaerder

    Energy Technology Data Exchange (ETDEWEB)

    Tavast, J. [ABB STAL AB, Finspaang (Sweden)

    1996-12-01

    Corrosion fatigue in LP steam turbine blading was reviewed together with result of tests performed in order to find blade materials with improved resistance against this. According to international experience, corrosion fatigue of 12Cr steam turbine blades in the transition zone between dry and wet steam, is one of the major causes, if not the major cause, for unavailability of steam turbines. Corrosion fatigue in LP blading is a frequent problem also in Swedish and Finnish nuclear power plants, especially in turbines of type D54 in BWR-plants. Corrosion fatigue has also been discovered in at least one type of nuclear turbine. Initiation times have been very long and the varying experiences in different types of turbines may simply reflect differing initiation times. Corrosion fatigue may therefore become more frequent in other types of turbines in the future. The type of water treatment (BWR/PWR) and possibly temperature after reheating seem to influence the risk for corrosion fatigue. Influence of inleakage of cooling water is less clear for these nuclear plants. The long initiation times together with the fact that very few of the cracked blades have actually failed, indicate that the cracks initiate and/or propagate during transients. Extensive laboratory tests show that there are alternative blade materials available with improved resistance against corrosion fatigue, with the most promising being 15/5 PH and A905, together with Ti6Al4V. The Ti alloy shows the best resistance against corrosion fatigue in most environments and is already used in some turbines. Disadvantage is a higher cost and possible need for redesign of the blades. The alternative materials are recommended for use for blades in the transition zone between dry and wet steam in LP turbines. The main disadvantage is a lack of references, even if 15%5 PH has been used to a very limited extent. 40 refs, 24 figs, 12 tabs, 9 appendices

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

    International Nuclear Information System (INIS)

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

  7. 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)

  8. Coalescing Wind Turbine Wakes

    Science.gov (United States)

    Lee, S.; Churchfield, M.; Sirnivas, S.; Moriarty, P.; Nielsen, F. G.; Skaare, B.; Byklum, E.

    2015-06-01

    A team of researchers from the National Renewable Energy Laboratory and Statoil used large-eddy simulations to numerically investigate the merging wakes from upstream offshore wind turbines. Merging wakes are typical phenomena in wind farm flows in which neighboring turbine wakes consolidate to form complex flow patterns that are as yet not well understood. In the present study, three 6-MW turbines in a row were subjected to a neutrally stable atmospheric boundary layer flow. As a result, the wake from the farthest upstream turbine conjoined the downstream wake, which significantly altered the subsequent velocity deficit structures, turbulence intensity, and the global meandering behavior. The complexity increased even more when the combined wakes from the two upstream turbines mixed with the wake generated by the last turbine, thereby forming a “triplet” structure. Although the influence of the wake generated by the first turbine decayed with downstream distance, the mutated wakes from the second turbine continued to influence the downstream wake. Two mirror-image angles of wind directions that yielded partial wakes impinging on the downstream turbines yielded asymmetric wake profiles that could be attributed to the changing flow directions in the rotor plane induced by the Coriolis force. The turbine wakes persisted for extended distances in the present study, which is a result of low aerodynamic surface roughness typically found in offshore conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  10. Detection of aeroacoustic sound sources on aircraft and wind turbines

    International Nuclear Information System (INIS)

    This thesis deals with the detection of aeroacoustic sound sources on aircraft and wind turbines using phased microphone arrays. First, the reliability of the array technique is assessed using airframe noise measurements in open and closed wind tunnels. It is demonstrated that quantitative acoustic measurements are possible in both wind tunnels. Then, the array technique is applied to characterize the noise sources on two modern large wind turbines. It is shown that practically all noise emitted to the ground is produced by the outer part of the blades during their downward movement. This asymmetric source pattern, which causes the typical swishing noise during the passage of the blades, can be explained by trailing edge noise directivity and convective amplification. Next, a semi-empirical prediction method is developed for the noise from large wind turbines. The prediction code is successfully validated against the experimental results, not only with regard to sound levels, spectra, and directivity, but also with regard to the noise source distribution in the rotor plane and the temporal variation in sound level (swish). The validated prediction method is then applied to calculate wind turbine noise footprints, which show that large swish amplitudes can occur even at large distance. The influence of airfoil shape on blade noise is investigated through acoustic wind tunnel tests on a series of wind turbine airfoils. Measurements are carried out at various wind speeds and angles of attack, with and without upstream turbulence and boundary layer tripping. The speed dependence, directivity, and tonal behaviour are determined for both trailing edge noise and inflow turbulence noise. Finally, two noise reduction concepts are tested on a large wind turbine: acoustically optimized airfoils and trailing edge serrations. Both blade modifications yield a significant trailing edge noise reduction at low frequencies, but also cause increased tip noise at high frequencies

  11. Wind Turbines Wake Aerodynamics

    DEFF Research Database (Denmark)

    Vermeer, L.; Sørensen, Jens Nørkær; Crespo, A.

    2003-01-01

    The aerodynamics of horizontal axis wind turbine wakes is studied. The contents is directed towards the physics of power extraction by wind turbines and reviews both the near and the far wake region. For the near wake, the survey is restricted to uniform, steady and parallel flow conditions......, thereby excluding wind shear, wind speed and rotor setting changes and yawed conditions. The emphasis is put on measurements in controlled conditions.For the far wake, the survey focusses on both single turbines and wind farm effects, and the experimental and numerical work are reviewed; the main interest...... is to study how the far wake decays downstream, in order to estimate the effect produced in downstream turbines.The article is further restricted to horizontal axis wind turbines and excludes all other types of turbines....

  12. Wind turbine wake aerodynamics

    Energy Technology Data Exchange (ETDEWEB)

    Vermeer, L.J. [Delft University of Technology (Netherlands). Section Wind Energy; Sorensen, J.N. [Technical University of Denmark, Lyngby (Denmark). Dept. of Mechanical Engineering; Crespo, A. [Universidad Politecnica de Madrid (Spain). Dpto. de Ingenieria Energetica y Fluidomecanica

    2003-10-01

    The aerodynamics of horizontal axis wind turbine wakes is studied. The contents is directed towards the physics of power extraction by wind turbines and reviews both the near and the far wake region. For the near wake, the survey is restricted to uniform, steady and parallel flow conditions, thereby excluding wind shear, wind speed and rotor setting changes and yawed conditions. The emphasis is put on measurements in controlled conditions. For the far wake, the survey focuses on both single turbines and wind farm effects, and the experimental and numerical work are reviewed; the main interest is to study how the far wake decays downstream, in order to estimate the effect produced in downstream turbines. The article is further restricted to horizontal axis wind turbines and excludes all other types of turbines. (author)

  13. Systematic comparison of trip distribution laws and models

    CERN Document Server

    Lenormand, Maxime; Ramasco, José J

    2016-01-01

    Trip distribution laws are basic for the travel demand characterization needed in transport and urban planning. Several approaches have been considered in the last years. One of them is the so-called gravity law, in which the number of trips is assumed to be related to the population at origin and destination and to decrease with the distance. The mathematical expression of this law resembles Newton's law of gravity, which explains its name. Another popular approach is inspired by the theory of intervening opportunities and it has been concreted into the so-called radiation models. Individuals are supposed to travel until they find a job opportunity, so the population and jobs spatial distributions naturally lead to a trip flow network. In this paper, we perform a thorough comparison between the gravity and the radiation approaches in their ability at estimating commuting flows. We test the gravity and the radiation laws against empirical trip data at different scales and coming from different countries. Diff...

  14. Geïntegreerde trips- en spintbestrijding in chrysant

    NARCIS (Netherlands)

    Beerling, E.A.M.; Berg, van den D.

    2008-01-01

    Raspportage van vierjarig project (2001-2004) waarbij op semi-praktijkschaal in proefkassen werd onderzocht wat de werkelijke bijdrage van natuurlijke vijanden en biologische middelen aan de bestrijding van trips en spint in chrysant is. Op basis hiervan werd geprobeerd zinvolle geïntegreerde bestri

  15. A model for sustainable short-term international medical trips.

    Science.gov (United States)

    Suchdev, Parminder; Ahrens, Kym; Click, Eleanor; Macklin, Lori; Evangelista, Doris; Graham, Elinor

    2007-01-01

    The health status of many people in developing countries is often dismal compared with the norms in industrialized countries. Increasingly, medical practitioners in the United States and other industrialized countries have become interested in global health issues, an interest that often takes the form of short-term international medical trips. We discuss several ethical issues associated with participation in such trips and use our experiences in developing the Children's Health International Medical Project of Seattle (CHIMPS) to outline and illustrate a set of 7 guiding principles for making these trips. CHIMPS is a resident-run, faculty-supported international medical program founded in 2002 by pediatric residents at the University of Washington in Seattle. Members of CHIMPS work with a rural community in El Salvador to support ongoing public health interventions there and provide sustainable medical care in collaboration with the community and a local nongovernmental organization. The 7 principles developed as a result of this work-mission, collaboration, education, service, teamwork, sustainability, and evaluation-can be used as a model for health practitioners as they develop or select international medical trips. The importance of partnering with the community and working within the existing medical and public health infrastructure is emphasized. Many of the challenges of doing international medical work can be overcome when efforts are guided by a few specific principles, such as those we have outlined. PMID:17660105

  16. At the Crossroads: Notes from a Trip to the Philippines

    OpenAIRE

    Schirmer, Daniel Boone

    1990-01-01

    Report of 1990 Philippines trip in preparation for the upcoming 1991 vote in the Philippine Senate to reject or renew the treaty between the Philippines and the United States regarding U.S. military bases. A conference, and subsequent travel around the country revealed the explosive growth of anti-bases sentiment.

  17. Solving touristic trip planning problem by using taboo search approach

    Directory of Open Access Journals (Sweden)

    Kadri Sylejmani

    2011-09-01

    Full Text Available In this paper, we introduce an algorithm that automatically plans a touristic trip by considering some hard and soft constrains. Opening and closing hours of POIs, trip duration and trip allocated budget represent the hard constraints, while the satisfaction factors of the POIs and travelling distance in the trip are considered as soft constraints. We use the soft constraints to evaluate the generated solution of the algorithm. The algorithm is developed by utilizing the taboo search method as a meta heuristic. The operators of Swap, Insert and Delete are used to explore the search space. The Swap and Insert operator are used in each iteration of the algorithm loop, while the Delete operator is used whenever the algorithm tends to enter in an endless cycle. The algorithm is developed by using Java programming language, while the data repositories are created in the XML format. The algorithm is tested with 40 instances of POIs of the city of Vienna. Various entry parameters of the algorithm are used to test its performance. The results gained are discussed and compared in respect to the optimal solution.

  18. The Scope of Gene Patent Protection and the TRIPS Agreement

    DEFF Research Database (Denmark)

    Sommer, Tine

    2007-01-01

    The Scope of Gene Patent Protection and the TRIPS Agreement - An Exclusively Nondiscriminatory Approach?   Gene patenting in Europe has provoked much debate both before and since the adoption of Directive 98/44/EC on the legal protection of biotechnological inventions. Some of the major points of...

  19. CNMI, American Samoa, and Guam Small Boat Fishery Trip Expenditure (2009 to present)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This is a time-series dataset of trip expenditure data including actual fishing trip expenses, input usage, and input prices, for boat-based reef fish, bottomfish,...

  20. High efficiency turbines

    OpenAIRE

    Varma, Vijaya Krushna Varma

    2012-01-01

    Varma designed ultra modern and high efficiency turbines which can use gas, steam or fuels as feed to produce electricity or mechanical work for wide range of usages and applications in industries or at work sites. Varma turbine engines can be used in all types of vehicles. These turbines can also be used in aircraft, ships, battle tanks, dredgers, mining equipment, earth moving machines etc,

  1. Wind Turbine Clutter

    OpenAIRE

    Gallardo-Hernando, Beatriz; Pérez-Martínez, Félix; Aguado-Encabo, Fernando

    2010-01-01

    In this chapter the main effects of wind turbines on the performance of radar systems have been explained. The radar signature of wind turbine clutter is unique and then, it requires a special treatment when developing mitigation techniques. WTC clutter remains spatially static, but it fluctuates continuously in time. In surveillance radars the return from wind turbines can be completely different from one scan to the following. In addition, apart from the powerful tower return, the movement ...

  2. Small wind turbine

    OpenAIRE

    Vélez Castellano, Didier

    2010-01-01

    The main objective is to develop a project on installing a small wind turbine at the University of Glyndwr in Wrexham Wales. Today are immersed in a world seeking clean energy for reduce greenhouse gases because this problem is becoming a global reality. So installing a small wind turbine at the university would provide large quantity of clean energy to supply a workshop and also reduce the expulsion of CO2 into the atmosphere. The main characteristic of the turbine under...

  3. Direct drive wind turbine

    Science.gov (United States)

    Bywaters, Garrett; Danforth, William; Bevington, Christopher; Stowell, Jesse; Costin, Daniel

    2006-07-11

    A wind turbine is provided that minimizes the size of the drive train and nacelle while maintaining the power electronics and transformer at the top of the tower. The turbine includes a direct drive generator having an integrated disk brake positioned radially inside the stator while minimizing the potential for contamination. The turbine further includes a means for mounting a transformer below the nacelle within the tower.

  4. A Heuristic for Locating Electric Vehicle Charging Stations for Trip Chains

    DEFF Research Database (Denmark)

    Wen, Min; Røpke, Stefan

    We present the problem of locating a limited number of electric vehiclecharging stations for a given set of trip chains, each of which consistsof a series of linked short trips and is represented by a sequence ofintervening stops along the trip chain. The objective of this problemis to maximize t...... number of trip chains that can be completed by the electric vehicle without running out of battery. A mixed-integer programmingformulation as well as a heuristic for solving this problemwill be presented....

  5. Rampressor Turbine Design

    Energy Technology Data Exchange (ETDEWEB)

    Ramgen Power Systems

    2003-09-30

    The design of a unique gas turbine engine is presented. The first Rampressor Turbine engine rig will be a configuration where the Rampressor rotor is integrated into an existing industrial gas turbine engine. The Rampressor rotor compresses air which is burned in a traditional stationary combustion system in order to increase the enthalpy of the compressed air. The combustion products are then expanded through a conventional gas turbine which provides both compressor and electrical power. This in turn produces shaft torque, which drives a generator to provide electricity. The design and the associated design process of such an engine are discussed in this report.

  6. Experimental turbine VT-400

    Science.gov (United States)

    Zitek, Pavel; Milčák, Petr; Noga, Tomáš

    2016-03-01

    The experimental air turbine VT400 is located in hall laboratories of the Department of Power System Engineering. It is a single-stage air turbine located in the suction of the compressor. It is able to solve various problems concerning the construction solution of turbine stages. The content of the article will deal mainly with the description of measurements on this turbine. The up-to-now research on this test rig will be briefly mentioned, too, as well as the description of the ongoing reconstruction.

  7. Experimental turbine VT-400

    OpenAIRE

    Zitek Pavel; Milčák Petr; Noga Tomáš

    2016-01-01

    The experimental air turbine VT400 is located in hall laboratories of the Department of Power System Engineering. It is a single-stage air turbine located in the suction of the compressor. It is able to solve various problems concerning the construction solution of turbine stages. The content of the article will deal mainly with the description of measurements on this turbine. The up-to-now research on this test rig will be briefly mentioned, too, as well as the description of the ongoing rec...

  8. Wind Turbine Structural Dynamics

    Science.gov (United States)

    Miller, D. R. (Editor)

    1978-01-01

    A workshop on wind turbine structural dynamics was held to review and document current United States work on the dynamic behavior of large wind turbines, primarily of the horizontal-axis type, and to identify and discuss other wind turbine configurations that may have lower cost and weight. Information was exchanged on the following topics: (1) Methods for calculating dynamic loads; (2) Aeroelasticity stability (3) Wind loads, both steady and transient; (4) Critical design conditions; (5) Drive train dynamics; and (6) Behavior of operating wind turbines.

  9. 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

  10. 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

  11. 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

  12. Cobra-TF simulation of BWR bundle dry out experiments

    International Nuclear Information System (INIS)

    The COBRA-TF computer code uses a two-fluid, three-field and three-dimensional formulation to model a two-phase flow field in a specific geometry. The liquid phase is divided in a continuous liquid field and a separate dispersed field, which is used to describe the entrained liquid drops. For each space dimension, the code solves three momentum equations, three mass conservation equations and two energy conservation equations. Entrainment and depositions models are implemented into the code to model the mass transfer between the two liquid fields. This study presents the results obtained with COBRA-TF for the simulation of the Siemens 9-9Q BWR Bundle Dryout experiments. The model includes 20 channels and 34 axial nodes in the heated section. The predicted critical power and dryout location is compared with the measured values. An assessment of the code entrainment and de-entrainment models is presented. (authors)

  13. 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

  14. 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)

  15. 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)

  16. 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)

  17. 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)

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

    International Nuclear Information System (INIS)

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

  19. 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

  20. 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)

  1. 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

  2. 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.)

  3. Fault tree analysis on BWR core spray system

    International Nuclear Information System (INIS)

    Fault Trees which describe the failure modes for the Core Spray System function in the Browns Ferry Nuclear Plant (BWR 1065MWe) were developed qualitatively and quantitatively. The unavailability for the Core Spray System was estimated to be 1.2 x 10-3/demand. It was found that the miscalibration of four reactor pressure sensors or the failure to open of the two inboard valves (FCV 75-25 and 75-53) could reduce system reliability significantly. It was recommended that the pressure sensors would be calibrated independently. The introduction of the redundant inboard valves could improve the system reliability. Thus this analysis method was verified useful for system analysis. The detailed test and maintenance manual and the informations on the control logic circuits of each active component are necessary for further analysis. (author)

  4. BWR Full Integral Simulation Test (FIST). Phase I test results

    International Nuclear Information System (INIS)

    A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report

  5. BWR Full Integral Simulation Test (FIST). Phase I test results

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, W S; Alamgir, M; Sutherland, W A

    1984-09-01

    A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report.

  6. 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

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

    International Nuclear Information System (INIS)

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

  8. 49 CFR 236.531 - Trip arm; height and distance from rail.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Trip arm; height and distance from rail. 236.531... Train Stop, Train Control and Cab Signal Systems Rules and Instructions; Roadway § 236.531 Trip arm; height and distance from rail. Trip arm of automatic train stop device when in the stop position shall...

  9. Questionnaire-based person trip visualization and its integration to quantitative measurements in Myanmar

    Science.gov (United States)

    Kimijiama, S.; Nagai, M.

    2016-06-01

    With telecommunication development in Myanmar, person trip survey is supposed to shift from conversational questionnaire to GPS survey. Integration of both historical questionnaire data to GPS survey and visualizing them are very important to evaluate chronological trip changes with socio-economic and environmental events. The objectives of this paper are to: (a) visualize questionnaire-based person trip data, (b) compare the errors between questionnaire and GPS data sets with respect to sex and age and (c) assess the trip behaviour in time-series. Totally, 345 individual respondents were selected through random stratification to assess person trip using a questionnaire and GPS survey for each. Conversion of trip information such as a destination from the questionnaires was conducted by using GIS. The results show that errors between the two data sets in the number of trips, total trip distance and total trip duration are 25.5%, 33.2% and 37.2%, respectively. The smaller errors are found among working-age females mainly employed with the project-related activities generated by foreign investment. The trip distant was yearly increased. The study concluded that visualization of questionnaire-based person trip data and integrating them to current quantitative measurements are very useful to explore historical trip changes and understand impacts from socio-economic events.

  10. Multiple Stages of Weekend Field Trips To Expose Students to Nature: Emphasis on Discovery and Awareness.

    Science.gov (United States)

    Krupa, James J.

    2002-01-01

    Describes the three stages of a field trip and reviews stage 1, weekend field trips, which focuses on an organism's morphology, behavior, and ecology. Presents activities on salamanders, small mammals, fish, birds, and bats. Explains the difficulties of weekend trips. (YDS)

  11. Nuevos atrayentes de trips ayudan a los agricultores en el control de plagas

    NARCIS (Netherlands)

    Tol, van R.W.H.M.; Kogel, de W.J.; Teulon, D.

    2007-01-01

    Los trips constituyen una plaga importante que afecta a muchos cultivos diferentes. El año pasado se probaron con éxito, en situaciones prácticas, aromas atrayentes de trips de las flores y trips de la cebolla. El producto, que estará a disposición de los cultivadores en junio, resultó efectivo en t

  12. Summary report of seismic PSA of BWR model plant

    International Nuclear Information System (INIS)

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

  13. Spectral effects in cavitation of BWR jet pumps

    International Nuclear Information System (INIS)

    BWR jet-pumps vibrate violently when the recirculation drive flow becomes sufficiently high that cavitation occurs in the throat section at the local temperature of the suction flow. This process occurs rapidly as the fluid flows through the nozzle throat and while mixing of the drive and suction flows is in process. Continued mixing during transit through the throat eventually results in a pressure recovery to the saturation pressure and subsequent collapse of the voids near the entrance to the diffuser. The liquid surrounding a vapor volume, or void, performs work during void collapse. The result is that energy is deposited in the liquid and converted locally to an over-pressure, or sonic wave. The power contained in this wave may be very large, since it is deposited in the relatively short void collapse time. The frequency spectrum of the sonic excitation depends on the temporal nature of the collapse, including multiple events when the void is composed of a swarm of bubbles of various sizes. The frequency spectrum of a single spherical void collapse is calculated from the bubble collapse dynamics as it propagates in an increasing ambient pressure field that is a function of drive flow rate, Qd. It is shown that the void collapse is a tandem process in the time frame of the bubble swarm. These considerations lead to comparisons with strain-gauge data taken in BWR tests with the head off. Averaged micro-strain data is compared with the calculated values at several pump speeds, then fit to the data set. The normalized result is in excellent agreement, indicating that the void collapse mechanism is the most likely source of vibration excitation in the jet pump structure

  14. Spectral effects in cavitation of BWR jet pumps

    Energy Technology Data Exchange (ETDEWEB)

    Terhune, J.H.; Karim-Panahi, K. [GE Nuclear Energy, San Jose, CA (United States)

    1996-12-01

    BWR jet-pumps vibrate violently when the recirculation drive flow becomes sufficiently high that cavitation occurs in the throat section at the local temperature of the suction flow. This process occurs rapidly as the fluid flows through the nozzle throat and while mixing of the drive and suction flows is in process. Continued mixing during transit through the throat eventually results in a pressure recovery to the saturation pressure and subsequent collapse of the voids near the entrance to the diffuser. The liquid surrounding a vapor volume, or void, performs work during void collapse. The result is that energy is deposited in the liquid and converted locally to an over-pressure, or sonic wave. The power contained in this wave may be very large, since it is deposited in the relatively short void collapse time. The frequency spectrum of the sonic excitation depends on the temporal nature of the collapse, including multiple events when the void is composed of a swarm of bubbles of various sizes. The frequency spectrum of a single spherical void collapse is calculated from the bubble collapse dynamics as it propagates in an increasing ambient pressure field that is a function of drive flow rate, Q{sub d}. It is shown that the void collapse is a tandem process in the time frame of the bubble swarm. These considerations lead to comparisons with strain-gauge data taken in BWR tests with the head off. Averaged micro-strain data is compared with the calculated values at several pump speeds, then fit to the data set. The normalized result is in excellent agreement, indicating that the void collapse mechanism is the most likely source of vibration excitation in the jet pump structure.

  15. Development of alternative materials for BWR fuel springs

    International Nuclear Information System (INIS)

    Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m2. Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58Co and 60Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  18. Using GIS for planning field trips: In-situ assessment of Geopoints for field trips with mobile devices

    Science.gov (United States)

    Böhm, Sarah; Kisser, Thomas; Ditter, Raimund

    2016-04-01

    Up to now no application is existing for collecting data via mobile devices using a geographical information system referring to the evaluation of Geopoints. Classified in different geographical topics a Geopark can be rated for suitability of Geopoints for field trips. The systematically acquisition of the suitability of Geopoints is necessary, especially when doing field trips with lower grade students who see a physical-geographic phenomenon for the first time. For this reason, the development of such an application is an invention for easy handling evaluations of Geopoints on the basis of commonly valid criteria like esthetic attraction, interestingness, and pithiness (Streifinger 2010). Collecting data provides the opportunity of receiving information of particularly suitable Geopoints out of the sight from students, tourists and others. One solution for collecting data in a simple and intuitive form is Survey123 for ArcGIS (http://survey123.esri.com/#/). You can create surveys using an ArcGIS Online organizational account and download your own survey or surveys "that may have been shared with you" (https://itunes.apple.com/us/app/survey-123-for-arcgis/id993015031?mt=8) on your mobile device. "Once a form is downloaded, you will be able to start collecting data."(https://itunes.apple.com/us/app/survey-123-for-arcgis/id993015031?mt=8) Free of cost and use while disconnected the application can easily be used via mobile device on field trips. On a 3-day field trip which is held three times per year in the Geopark Bergstraße-Odenwald Survey123 is being used to evaluate the suitability of different Geopoints for different topics (geology, soils, vegetation, climate). With every field trip about 25 students take part in the survey and evaluate each Geopoint at the route. So, over the time, the docents know exactly which Geopoints suites perfect for teaching geology for example, and why it suites that good. The field trip is organized in an innovative way. Before

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

    International Nuclear Information System (INIS)

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

  20. Design and optimization of HPLWR high pressure Turbine gamma ray shield

    International Nuclear Information System (INIS)

    Highlights: • Shield around HPLWR high pressure turbine optimized by Monte Carlo method. • The occupancy period in the turbine building was considered in the optimization. • Shield thickness is significantly reduced when heavy concretes are used. • Shield thickness for BWRs is sufficient for HPWR if heavy concrete is used. - Abstract: This work proposes the optimum gamma ray shield thickness around the HPLWR high pressure turbine for different occupancy periods in the turbine building. Monte Carlo method was employed in the design process and only radioactive nitrogen-16 was considered as the source of radiation. Five grades of concrete (ordinary, magnetite, heavy magnetite, steel magnetite and barite) were used as shielding materials. The isotope source term in the high pressure turbine was estimated by modeling the HPLWR three pass core in MCNP and tracking the inventory using a simple algorithm. The high pressure turbine was thereafter modeled in MCNP with a concrete shield arrayed in layers around it. The surface flux tally and ICRP74 dose conversion coefficients were employed to estimate the dose profile across the shield. For some shielding materials, exponential functions were fitted on the calculated data to extrapolate dose values beyond the model thickness. The optimum shield thickness was determined by comparing the calculated dose profiles with dose limit proposals in the IAEA standard (NS-G-1.13) on radiation protection considerations during nuclear power plant design. It was observed that with a 120 cm thick heavy concrete shield, the turbine building would be safe for most occupancy periods. However for ordinary concrete the shield would require some extension to guarantee safety. For very long occupancy (more than 10 person hours per week), magnetite shield may also require slight extension. It can therefore be concluded that the shield thickness recommended for BWR turbines (which operate on a direct cycle like HPLWR) could be sufficient

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

    International Nuclear Information System (INIS)

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

  2. An Engineering Process for Automatic Seismic Trip System Implementation

    International Nuclear Information System (INIS)

    To cope with potential seismic risks of catastrophic earthquakes, an Automatic Seismic Trip System (ASTS) is being developed for the operating nuclear power plants in Korea. The ASTS is designed to automatically trip the reactor upon an occurrence of the Safe Shutdown Earthquake (SSE) at the plant site. By the end of 2012, the ASTS will have been installed at all twenty (20) nuclear power plants currently under operation in Korea as of 2010. The system must be designed and constructed so that it not only ensures a highly reliable reactor trip upon the SSE but also minimizes a spurious trip during normal operation and maintenance. Since the ASTS is designed for the currently operating plants, the system design must consider different reactor types, including the Pressurized Water Reactor (PWR) and the Pressurized Heavy Water Reactor (PHWR), and easy operation and maintenance as well. This paper presents an engineering process for the design and implementation of the ASTS for the nuclear power plants. The presentation is mainly focused on the application of a systematic design process and a rigorous verification process for the various nuclear power plants. The systematic design process is based on the concepts of modularization and standardization. For the systematic design, the ASTS is functionally divided into three separate modules consisting of a sensor module, a trip logic module, and a trip actuation module. Rigorous verification is applied to the hardware qualification and the software verification. Environmental, seismic, and EMI qualifications are included in hardware verification. Strict software verification is also performed through the entire life cycle for software development. Even though the ASTS is not designed as a safety-related Class 1E system per the requirement of ANS-51.1, the intent of safety related design standards are applied, to the extent practical, to the hardware qualification and software verification. Application of this engineering

  3. Improvement of turbine materials

    International Nuclear Information System (INIS)

    Materials for turbine blades and rotors are discussed with a view to the following subjects: Long period creep behaviour, gas/metal reactions, fatigue behaviour in long-term and creep strength testing, fracture mechanics testing, creep/fatigue interactions, development of a turbine blade of TZM, jointing of TZM, decontamination. (orig./IHOE)

  4. Gas turbine diagnostic system

    CERN Document Server

    Talgat, Shuvatov

    2011-01-01

    In the given article the methods of parametric diagnostics of gas turbine based on fuzzy logic is proposed. The diagnostic map of interconnection between some parts of turbine and changes of corresponding parameters has been developed. Also we have created model to define the efficiency of the compressor using fuzzy logic algorithms.

  5. Wind turbine state estimation

    DEFF Research Database (Denmark)

    Knudsen, Torben

    2014-01-01

    which was successful. Then the estimation of a wind turbine state including dynamic inflow was tested on a simulated NREL 5MW turbine was performed. This worked perfectly with wind speeds from low to nominal wind speed as the output prediction errors where white. In high wind where the pitch actuator...

  6. Design optimization and modification of reactor trip protection system of nuclear power unit of WWER

    International Nuclear Information System (INIS)

    When carrying out periodic test of nuclear power unit reactor trip protection system, it is necessary to disconnect trip breakers in turn. There are big risks in such kind of periodic tests and many unplanned trip events happened because of malfunction of equipment in reactor trip protection system periodic tests of foreign and domestic nuclear power units under operation, which caused great economic loss. The paper introduces the practical results of design optimization and modification of reactor trip protection system of a nuclear power unit of WWER. (authors)

  7. Nuclear turbine power plant

    International Nuclear Information System (INIS)

    Purpose : To improve the heat cycle balance in a nuclear turbine power plant or the like equipped with a moisture separating and reheating device, by eliminating undesired overcooling of the drains in the pipes of a heat transmission pipe bundle. Constitution : A high pressure turbine is driven by main steams from a steam generator. The steams after driving the high pressure turbine are removed with moistures by way of a moisture separator and then re-heated. Extracted steams from the steam generator or the high pressure turbine are used as a heating source for the reheating. In the nuclear turbine power plant having such a constitution, a vessel for separating the drains and the steams resulted from the heat exchange is provided at the outlet of the reheating device and the steams in the vessel are introduced to the inlet of the moisture separator. (Aizawa, K.)

  8. Noise from wind turbines

    International Nuclear Information System (INIS)

    Denmark has 3200 wind turbines with an installed maximum capacity of 418MW. The most important Danish research projects into wind turbine noise and the main results are listed. These date from 1983. Two comprehensive studies are currently in progress. The first is an analytical and empirical investigation of aerodynamic noise from wind turbine rotors and has so far dealt mainly with tip noise. The measurement method, using a hard board mounted microphone on the ground near the turbine, is described. Four different tip designs have been tested. Some examples of reference sound power level spectra for three of the designs are presented. During the past two years a computerbased data acquisition system has been used for real-time determination of sound power levels. The second study, which has just commenced, is on annoyance from wind turbine noise. It will include noise measurements, masking calculations and a social survey on the perceived nuisance. (UK)

  9. Ceramic Cerami Turbine Nozzle

    Science.gov (United States)

    Boyd, Gary L.

    1997-04-01

    A turbine nozzle vane assembly having a preestablished rate of thermal expansion is positioned in a gas turbine engine and being attached to conventional metallic components. The metallic components having a preestablished rate of thermal expansion being greater than the preestablished rate of thermal expansion of the turbine nozzle vane assembly. The turbine nozzle vane assembly includes an outer shroud and an inner shroud having a plurality of horizontally segmented vanes therebetween being positioned by a connecting member positioning segmented vanes in functional relationship one to another. The turbine nozzle vane assembly provides an economical, reliable and effective ceramic component having a preestablished rate of thermal expansion being greater than the preestablished rate of thermal expansion of the other component.

  10. Ceramic turbine nozzle

    Science.gov (United States)

    Shaffer, James E.; Norton, Paul F.

    1996-01-01

    A turbine nozzle and shroud assembly having a preestablished rate of thermal expansion is positioned in a gas turbine engine and being attached to conventional metallic components. The metallic components having a preestablished rate of thermal expansion being greater than the preestablished rate of thermal expansion of the turbine nozzle vane assembly. The turbine nozzle vane assembly includes a plurality of segmented vane defining a first vane segment and a second vane segment. Each of the first and second vane segments having a vertical portion. Each of the first vane segments and the second vane segments being positioned in functional relationship one to another within a recess formed within an outer shroud and an inner shroud. The turbine nozzle and shroud assembly provides an economical, reliable and effective ceramic component having a preestablished rate of thermal expansion being less than the preestablished rate of thermal expansion of the other component.

  11. Using a Field Trip Inventory to Determine If Listening to Elementary School Students' Conversations, While on a Zoo Field Trip, Enhances Preservice Teachers' Abilities to Plan Zoo Field Trips

    Science.gov (United States)

    Patrick, Patricia; Mathews, Cathy; Tunnicliffe, Sue Dale

    2013-01-01

    This study investigated whether listening to spontaneous conversations of elementary students and their teachers/chaperones, while they were visiting a zoo, affected preservice elementary teachers' conceptions about planning a field trip to the zoo. One hundred five preservice elementary teachers designed field trips prior to and after…

  12. Babcock and Wilcox Owners' Group program: Trip reduction and transient response improvement

    International Nuclear Information System (INIS)

    In 1985, the average trip frequency for the industry was 4.3 trips per plant per year while Babcock ampersand Wilcox (B ampersand W)-designed plants had 4.5 trips. In early 1986, the B ampersand W Owners' Group (B ampersand WOG) established goals to reduce trip frequency and improve posttrip transient response. Through the recommendations of the B ampersand WOG Trip Reduction and Transient Response Improvement Program (TR/TRIP) and other utility initiatives, the trip frequency for the B ampersand WOG plants has been on a progressive downward trend and has been consistently below the industry average since 1986. The successful results in trip reduction for the B ampersand WOG plants are shown. The B ampersand WOG has implemented several programs that have resulted in fewer trips per plant. This success can be attributed to the following: (1) a comprehensive program to evaluate each trip and transient for root-cause determination, define corrective actions, share information, and peer reviews; (2) a broad program to review systems and components that contribute to trips and transients, identify specific recommendations to correct deficiencies, utility commitment to implementation, conduct internal monitoring and indirectly exert peer pressure; (3) an awareness of the goals at all levels in the organization coupled with strong executive-level involvement; and (4) timely implementation of recommendations

  13. The Big Trip and Wheeler-DeWitt equation

    CERN Document Server

    Yurov, Artyom V; Yurov, Valerian A

    2012-01-01

    Of all the possible ways to describe the behavior of the universe that has undergone a big trip the Wheeler-DeWitt equation should be the most accurate -- provided, of course, that we employ the correct formulation. In this article we start by discussing the standard formulation introduced by Gonz\\'alez-D\\'iaz and Jimenez-Madrid, and show that it allows for a simple yet efficient method of the solution's generation, which is based on the Moutard transformation. Next, by shedding the unnecessary restrictions, imposed on aforementioned standard formulation we introduce a more general form of the Wheeler-DeWitt equation. One immediate prediction of this new formula is that for the universe the probability to emerge right after the big trip in a state with $w=w_0$ will be maximal if and only if $w_0=-1/3$.

  14. 10 years after implementation of TRIPS obligations in India.

    Science.gov (United States)

    Pawar, Mahesh D

    2015-01-01

    The Patent (Amendment) Act of 2005 enforced after TRIPS raises many issues which hinder growth of Indian pharma companies. To tackle this, Indian pharma companies doubled their R&D expenditure and became significant players in global generic drug market. Indian pharmaceutical companies, which predominantly focused on import-oriented market, shifted to research-based approach by signing various agreements with MNCs that led to M&A and technology transfer. At the same time growth in R&D activities increased ANDA and DMF filing in the USA and Europe. Companies also kept their social responsibility by selling medicines at affordable price to patients. This paper highlights the changing business scenario of Indian pharmaceutical companies to counteract various issues evolved from new patent regime after TRIPS. PMID:26030078

  15. Analysis of reactor trips originating in balance of plant systems

    Energy Technology Data Exchange (ETDEWEB)

    Stetson, F.T.; Gallagher, D.W.; Le, P.T.; Ebert, M.W. (Science Applications International Corp., McLean, VA (USA))

    1990-09-01

    This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs.

  16. Microstructure Evolution during Friction Stir Spot Welding of TRIP steel

    DEFF Research Database (Denmark)

    Lomholt, Trine Colding

    Transformation Induced Plasticity (TRIP) steels have been developed for automotive applications due to the excellent high strength and formability. The microstructure of TRIP steels is a complex mixture of various microstructural constituents; ferrite, bainite, martensite and retained austenite...... deteriorating the uniform elongation. The unique deformation properties can be exploited in automotive applications for crash resistant parts due to the high energy absorption, thus improving passenger safety. Furthermore, the high strength and good formability permits the application of thinner sheet material...... Stir Spot Welding (FSSW) is investigated. The aim of the study is to assess whether high quality welds can be produced and, in particular, to obtain an understanding of the microstructural changes during welding. The microstructure of the welded samples was investigated by means of reflected light...

  17. Analysis of reactor trips originating in balance of plant systems

    International Nuclear Information System (INIS)

    This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs

  18. Visiting the USSR: a trip of a lifetime.

    Science.gov (United States)

    Brown, Geraldine

    2012-01-01

    To Russia, who me? That is actually how it all began. A decade or more ago, I had the opportunity to visit what was then known as the Union of Soviet Socialist Republics (USSR), Russia. Although, this place held high priority on my list of places to go, I never thought such a trip was within my reach. This idea was quite fascinating to me because of the events that did happen there, including the Russian space ship Sputnik and the dog, the high stepping military officers, Red Square and the St. Basil's Cathedral. After reading a lot about Russia, I thought it would be great to see a clean place, where it was unlawful to throw paper on the streets, and ride in public transportation such as the buses, taxis and the subway system, which were immaculately clean. It was an exciting trip, one, I will always remember, but would be a difficult adjustment to make, to live. PMID:22774358

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

    International Nuclear Information System (INIS)

    Full text: The innovative concept for a reduced-moderation spectrum boiling water reactor (BWR) combined with an advanced fuel cycle system, which is named BARS, has been being developed as one of the IVNET (Innovative and Viable Nuclear Energy Technology) development projects from 2000 in Japan. The reduced-moderation spectrum BWR with high conversion property features higher utilization of uranium resources, multi-recycling usage of plutonium and flexibility in loading the extracted long-lived radioisotopes, it can realize a sustainable nuclear energy system without a liquid metal fast breeder reactor (LMFBR). The advanced recycle system means the combination of dry reprocessing and vibro-packing of MOX fuel fabrication. This system simplifies reprocessing and MOX fuel fabricating process and reduces related back end cost. The reduced moderation spectrum condition in BWR is obtained through triangular tight fuel rod lattice configuration and higher void fraction. This feature cause some technological problems; smaller void coefficients of positive side and larger uncertainty of nuclear and thermal hydraulic design method. According to the above problems, the developing project consists the three parts; core design study that enables the negative void coefficient, critical assembly experiment to validate nuclear design method and thermal hydraulic test to evaluate thermal hydraulic property. The core design study focuses on optimization of tight lattice bundle design. In the optimization of bundle design, it is important to satisfy both of the requirements for high conversion ratio and negative void coefficient. Streaming channel concept has been introduced to realize the negative void coefficient. The streaming channel consists of short bundle and upper vacant tube, which promotes neutron-streaming effect in higher void fraction. The neutron-streaming effect evaluated by Monte Carlo calculation showed the effectiveness of streaming channel concept for realizing

  20. 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

  1. Prediction of the mechanical behaviour of TRIP steel

    OpenAIRE

    Perdahcioglu, E.S.; Geijselaers, H.J.M.; Tekkaya, A. E.; Hirt, G.

    2011-01-01

    TRIP steel typically contains four different phases, ferrite, bainite, austenite and martensite. During deformation the metastable retained austenite tends to transform to stable martensite. The accompanying transformation strain has a beneficial effect on the ductility of the steel during forming. By changing the alloy composition, the rolling procedure and the thermal processing of the steel, a wide range of different morphologies and microstructures can be obtained. Interesting parameters ...

  2. Are short daily trips compensated by higher leisure mobility?

    OpenAIRE

    Petter Næss

    2006-01-01

    Studies in several cities have shown that inner-city residents travel shorter distances and use cars less for local transport than suburbanites do. However, according to some authors, a low daily amount of travel is likely to be compensated through more extensive leisure mobility at weekends and on holidays. On the basis of a study of residential location and travel in the Copenhagen metropolitan area, this paper addresses the phenomenon of compensatory travel. For travel within ‘weekend trip...

  3. EFFECT OF CHEMICAL COMPOSITION ON RETAINED AUSTENITE IN TRIP STEEL

    Institute of Scientific and Technical Information of China (English)

    Y. Chen; X. Chen; Q.F. Wang; G.L. Yuan; C.Y. Li; X.Y. Li; Y.X. Wang

    2002-01-01

    The systematic chemical compositions including common C, Si, Mn, Al, and micro- alloying elements of Ti and Nb were designed for high volume fraction of retained austenite as much as possible. The thermo-cycle experiments were conducted by using Gleeble 2000 thermo-dynamic test machine for finding the appropriate composition. The experimental results showed that chemical composition had a significant effect on retained austenite, and the appropriate compositions were determined for commercial production of TRIP steels.

  4. Marketing a destination: Case of CreateTrips and Mexico

    OpenAIRE

    Tiainen, Johanna; Korvenpää, Emmi

    2015-01-01

    This thesis concentrates on Finnish people travelling to Mexico. Firstly, the writers conduct a quantitative research, a questionnaire, that studies Finnish people’s thoughts and presumptions about Mexico. Secondly, they create mobile travel guides of four different destinations. The questionnaire concentrates on the people’s point of view, asking what people think about Mexico, on what kind of trip would they go it they travel there, how long it would last and so on. The questionnaire also h...

  5. The Use of trips Flexibilities by Emerging Countries.

    OpenAIRE

    Amélie Robine

    2007-01-01

    The patent law has been harmonized at the international level in 1994 by the Agreement on Trade- Related Aspects of Intellectual Property Rights (trips). In practice, this text requires all member States of the World Trade Organization (wto) to recognize the patentability in all industrial sectors. A pharmaceutical patent shall then be granted for any pharmaceutical invention insofar as it fulfills all three criteria for patentability: novelty, inventive step and industrial application. Thus,...

  6. Meaningful Field Trip in Education of Renewable Energy Technologies

    OpenAIRE

    Hasan Said Tortop

    2013-01-01

    Renewable energy sources, in terms of countries‟ obtaining their energy needs from clean and without harming the environment is becoming increasingly important. This situation also requires improving the quality of science education will be given in this field. In this activity, in a field trip to the center for the renewable energy resources technologies, the application of learning cycle model appropriate for constructivist approach is shown. In the example of solar chimney activity accordi...

  7. Equilibrium analysis of trip chains in congested networks

    OpenAIRE

    Heydecker, B.

    2004-01-01

    In this paper, we develop a model of travel in a chain of trips joining several locations througha congested network. We develop a microscopic analysis of individual benefits obtained byspending time at each of the locations and costs incurred through travel between them. This iscombined with a macroscopic equilibrium model of travel during congested peak periods toshow how individuals? travel choices are influenced by the congestion that result fromcorresponding choices made by others. We sh...

  8. Bicycle-Sharing System Analysis and Trip Prediction

    OpenAIRE

    Zhang, Jiawei; Pan, Xiao; Li, Moyin; Yu, Philip S.

    2016-01-01

    Bicycle-sharing systems, which can provide shared bike usage services for the public, have been launched in many big cities. In bicycle-sharing systems, people can borrow and return bikes at any stations in the service region very conveniently. Therefore, bicycle-sharing systems are normally used as a short-distance trip supplement for private vehicles as well as regular public transportation. Meanwhile, for stations located at different places in the service region, the bike usages can be qu...

  9. Understanding Social Learning Behaviors via a Virtual Field Trip

    Directory of Open Access Journals (Sweden)

    Xin Bai

    2014-06-01

    Full Text Available This is a multidisciplinary study investigating how a virtual rather than face-to-face field trip can be conducted in a real-world setting and how students respond to such a social learning opportunity. Our participants followed a story of a stroke patient at her virtual home and in a virtual hospital via a teaching vignette. They were then given a new case and got on a virtual trip via a multiuser virtual environment. They played the roles of patients, relatives, doctors, or nurses, experiencing the emotional, physical, or social impacts those stakeholders may go through. Our study finds the overall participation of the Virtual Group is 50% more than the Text Group. Although the Virtual Group generates much more nodes in total, they focused much less on knowledge sharing and comparing than the Text Group (46 vs. 67, but more on other higher-level aspects of social interactions, such as knowledge discovery (57 vs. 42, co-construction (66 vs. 39, testing and modification (58 vs. 24 and application of newly constructed meaning (60 vs. 16. Analysis of students’ virtual field activities and in-depth discussions of important issues implied are included to help understand social learning behaviors during a virtual field trip. Sustainability of such systems is discussed.

  10. Controlling the feedwater flow in a BWR. Examples from Forsmark 2

    International Nuclear Information System (INIS)

    An investigation of the feedwater controller at Forsmark 2 has been performed. The investigation is based on signal analysis of measurement signals recorded during operation of the plant during different tests. The feedwater controller consists of the water level controller, the flow controller and the condenser balance controller. The overall goal of the feedwater control is to maintain constant water level (level controller) in the reactor and at the same time balance the water levels in the two condensers (condenser balance controller) to avoid that one condenser is full of water while the other one is operated with too low level. There is also a feed forward of the difference between steam flow and feedwater flow (flow controller) for each turbine system with the aim to reduce the fluctuation in reactor water level. The relation in strength between the three controllers is such that the level controller is the strongest followed by the condenser balance controller and finally the flow controller. Tests with trip of the feedwater pump and automatic start of the spare pump in each turbine system indicates a fast reduction in reactor water level that is restored after the transient in the control system. The transient in water level is stable without oscillations. However, it takes about 100 s before the reactor water level is restored. The function of the flow controller has been questioned by the authors. It does not take the action that is expected when a disturbance takes place in the difference between steam and feedwater flow. In addition to this principal weakness there is an offset in the feedwater controller output for feedwater flow 22 that reduces the contribution in flow control that is expected during the introduction of a disturbance. This offset should be adjusted during instrument maintenance of the feedwater controller. The PIP parameters for the level controller are gain factors and time constants. These have been evaluated with the aid of

  11. Properties and application of TRIP-steel in sheet metal forming

    Energy Technology Data Exchange (ETDEWEB)

    Doege, E.; Kulp, S.; Sunderkoetter, C. [Dept. of Sheet Metal Forming, Inst. for Metal Forming and Metal Forming Machine Tools, Univ. of Hannover (Germany)

    2002-07-01

    A further development of dual-phase-steels are represented by TRIP (transformation induced plasticity) -steels. TRIP-steels contain austenite, which is metastable at room temperature. It transforms to martensite during straining (TRIP effect). This process improves the strength-ductility balance of these steels. Two types of TRIP-steels, low alloyed (L-TRIP) and high alloyed (H-TRIP), can be applied in sheet forming processes and exhibit different forming characteristics. Basing on results of uniaxial tensile tests and the evaluation of Young's modulus the forming limits in deep drawing processes and the component properties of deep drawn parts are discussed. The Young's modulus decreases significantly with increasing pre-strain, especially demonstrated for the L-TRIP material TRIP700. Forming limit curves determined at different forming temperatures indicate its influence on the forming limits. Martensite transformation is suppressed at a temperature of approximately T = 200 C and therefore the major strain {phi}{sub 1} decreases significantly. For the investigated stainless steel AISI304 (H-TRIP) different lubricant types in comparison to chlorinated paraffins have been tested. Lubricants consisting of sulphur additives led to good forming conditions in forming processes, even better than lubricants based on chlorinated paraffins. The evaluation of component properties, compared between L-TRIP and H-TRIP, was done based on the analysis of springback and dent resistance. The L-TRIP material TRIP700 shows higher springback angles than AISI304 resulting from higher yield strength and decreased Young's modulus, resulting from the forming process. The dent resistance of TRIP-steel was exemplarily demonstrated for AISI304. Uniaxial pre-strained sheet specimen were analysed to show the dent resistance depending on dent depth. During elastic denting prestrain has no influence on dent resistance. Further increasing dent depth lead to increased dent forces

  12. Functional characterization of Trip10 in cancer cell growth and survival

    Directory of Open Access Journals (Sweden)

    Yan Pearlly S

    2011-02-01

    Full Text Available Abstract Background The Cdc42-interacting protein-4, Trip10 (also known as CIP4, is a multi-domain adaptor protein involved in diverse cellular processes, which functions in a tissue-specific and cell lineage-specific manner. We previously found that Trip10 is highly expressed in estrogen receptor-expressing (ER+ breast cancer cells. Estrogen receptor depletion reduced Trip10 expression by progressively increasing DNA methylation. We hypothesized that Trip10 functions as a tumor suppressor and may be involved in the malignancy of ER-negative (ER- breast cancer. To test this hypothesis and evaluate whether Trip10 is epigenetically regulated by DNA methylation in other cancers, we evaluated DNA methylation of Trip10 in liver cancer, brain tumor, ovarian cancer, and breast cancer. Methods We applied methylation-specific polymerase chain reaction and bisulfite sequencing to determine the DNA methylation of Trip10 in various cancer cell lines and tumor specimens. We also overexpressed Trip10 to observe its effect on colony formation and in vivo tumorigenesis. Results We found that Trip10 is hypermethylated in brain tumor and breast cancer, but hypomethylated in liver cancer. Overexpressed Trip10 was associated with endogenous Cdc42 and huntingtin in IMR-32 brain tumor cells and CP70 ovarian cancer cells. However, overexpression of Trip10 promoted colony formation in IMR-32 cells and tumorigenesis in mice inoculated with IMR-32 cells, whereas overexpressed Trip10 substantially suppressed colony formation in CP70 cells and tumorigenesis in mice inoculated with CP70 cells. Conclusions Trip10 regulates cancer cell growth and death in a cancer type-specific manner. Differential DNA methylation of Trip10 can either promote cell survival or cell death in a cell type-dependent manner.

  13. Cooled snubber structure for turbine blades

    Science.gov (United States)

    Mayer, Clinton A; Campbell, Christian X; Whalley, Andrew; Marra, John J

    2014-04-01

    A turbine blade assembly in a turbine engine. The turbine blade assembly includes a turbine blade and a first snubber structure. The turbine blade includes an internal cooling passage containing cooling air. The first snubber structure extends outwardly from a sidewall of the turbine blade and includes a hollow interior portion that receives cooling air from the internal cooling passage of the turbine blade.

  14. 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

  15. Pre-trip vs. post-trip destination image variations: A case of inbound tourists to Tanzania

    OpenAIRE

    Jani, Dev; Nguni, Winnie

    2016-01-01

    Despite the conceptual indications of destination image to vary with factors like travel status and tourist behaviour, there are few studies that have empirically researched the variations in destination image particularly in emerging destinations like those in Africa. This study aimed at testing the variation of pre- and post trip destination image held by inbound travellers to Tanzania. Four hypotheses were defined to test variation of destination image with travel status, destination famil...

  16. Composite turbine bucket assembly

    Science.gov (United States)

    Liotta, Gary Charles; Garcia-Crespo, Andres

    2014-05-20

    A composite turbine blade assembly includes a ceramic blade including an airfoil portion, a shank portion and an attachment portion; and a transition assembly adapted to attach the ceramic blade to a turbine disk or rotor, the transition assembly including first and second transition components clamped together, trapping said ceramic airfoil therebetween. Interior surfaces of the first and second transition portions are formed to mate with the shank portion and the attachment portion of the ceramic blade, and exterior surfaces of said first and second transition components are formed to include an attachment feature enabling the transition assembly to be attached to the turbine rotor or disk.

  17. RETRAN analysis results of feedwater pump trip transient for Lungmen ABWR Plant

    International Nuclear Information System (INIS)

    Highlights: → The RETRAN model was used to predict one feedwater pump trip (FWPT) transient. → The result shows that the margin sustains at least 30 cm above the L3 setpoint. → The unavailable motor driven pump case eventually actuates the low level scram signal. → The lowest load line case without motor driven pump still actuates the L3 scram. - Abstract: The RETRAN model of Lungmen ABWR was used to simulate one feedwater pump trip (FWPT) transient of the Lungmen start-up test program. The purpose of this test is to verify the capability of one surviving Turbine Driven Reactor Feedwater Pump (TDRFP) plus a Motor Driven Feedwater Pump (MDRFP) to continue operating the reactor stably following the incident. There are three major control systems implanted in Lungmen RETRAN model (LRM), which include Recirculation Flow Control System (RFCS), Steam Bypass and Pressure Control System (SBPCS), and Feedwater Control System (FWCS). The reactor water level margin with respect to the low level scram setpoint in the transient is monitored to confirm whether the acceptance criteria has been satisfied, which depends on the responses of the control systems to the FWPT transient. The analysis result of base case at 100% rated power/100% rated core flow with automatic start of MDRFP demonstrates that the acceptance criteria are met, which shows that the water level still sustains ample margin of 30 cm above the low level setpoint, and the reactor does not scram. To get more insight into the function of MDRFP, a set of sensitivity studies with the assumption of unavailable MDRFP, and with a different initial condition which extended to the maximum allowable core flow of 111% rated at rated power, was conducted to verify the superior capability of power coastdown due to the RIPs runback logic under the lowest load line, and also the delay time of the Reactor Internal Pumps (RIPs). Finally, it is concluded that FWPT transient without start of MDRFP eventually actuates the low

  18. Spreading Geodiversity awareness in schools through field trips and ICT

    Science.gov (United States)

    Magagna, Alessandra; Giardino, Marco; Ferrero, Elena

    2014-05-01

    Geodiversity, unlike Biodiversity, is not a topic included in the Italian schools curriculum. Nevertheless, Geomorphology is taught at all levels, and it seems to be the right tool for introducing the students to the concepts related to Geodiversity. In this context, a research on the use of field trips and Information and Communication Technologies (ICT) is being carried out for spreading the value of Geodiversity in Secondary Schools. Relevant international literature states that field trips are effective didactic tools for Earth Science education, because they stimulate an active learning process and allow students to appreciate the geological complexity of an area. On the other side, ICT allow students to get knowledge about the variety of landforms of their own territory by staying indoor, using virtual field trips and free software like Google Earth, Google Maps, Bing etc. In order to connect the two strategies, an innovative educational project is proposed here; it involves both the indoor and the outdoor activities, by enhancing a critical approach to the complexity of geological processes. As a starting point, a multimedia product on 20 Italian geological tours, designed for analyzing Geodiversity at a regional scale, has been tested with teachers and students, in order to understand its effectiveness by using it solely indoor. In a second phase, teachers and students have been proposed to compare and integrate indoor and outdoor activities to approach Geodiversity directly at a local scale, by means of targeted field trips. For achieving this goal, during the field trips, students used their mobile devices (smartphone and tablet) equipped with free and/or open source applications (Epicollect, Trimble Outdoor Navigator). These tools allow to track field trips, to gather data (geomorphological observations and related photographs), and to elaborate them in the laboratory; a process useful for reasoning on concepts such as spatial and temporal scales and for

  19. Predictive Process Optimization for Fracture Ductility in Automotive TRIP Steels

    Science.gov (United States)

    Gong, Jiadong

    In light of the emerging challenges in the automotive industry of meeting new energy-saving and environment-friendly requirements imposed by both the government and the society, the auto makers have been working relentlessly to reduce the weight of automobiles. While steel makers pushed out a variety of novel Advanced High Strength Steels (AHSS) to serve this market with new needs, TRIP (Transformation Induced Plasticity) steels is one of the most promising materials for auto-body due to its exceptional combination of strength and formability. However, current commercial automotive TRIP steels demonstrate relatively low hole-expansion (HE) capability, which is critical in stretch forming of various auto parts. This shortcoming on ductility has been causing fracture issues in the forming process and limits the wider applications of this steel. The kinetic theory of martensitic transformations and associated transformation plasticity is applied to the optimization of transformation stability for enhanced mechanical properties in a class of high strength galvannealed TRIP steel. This research leverages newly developed characterization and simulation capabilities, supporting computational design of high-performance steels exploiting optimized transformation plasticity for desired mechanical behaviors, especially for the hole-expansion ductility. The microstructure of the automotive TRIP sheet steels was investigated, using advanced tomographic characterization including nanoscale Local Electrode Atom Probe (LEAP) microanalysis. The microstructural basis of austenite stability, the austenite carbon concentration in particular, was quantified and correlated with measured fracture ductility through transformation plasticity constitutive laws. Plastic flow stability for enhanced local fracture ductility at high strength is sought to maintain high hole-expansion ductility, through quantifying the optimal stability and the heat-treatment process to achieve it. An additional

  20. Small break LOCA analysis for RCP trip strategy for YGN 3 and 4 emergency procedure guidelines

    International Nuclear Information System (INIS)

    A continued operation of RCPs during a certain small break LOCA may increase unnecessary inventory loss from the RCS causing a severe core uncovery which might lead to a fuel failure. After TMI-2 accident, the CEOG developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2) in response to NRC requests and incorporated it in the generic EPG for CE plants. The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis determines the RCP trip setpoint and demonstrates the safe operational aspects of RCP trip strategy during a small break LOCA for YGN 3 and 4. The trip setpoint of the first two RCPs for YGN 3 and 4 is calculated to be 1775 psia in pressurizer pressure based on the limiting small break LOCA with 0.15 ft2 break size in the hot leg. The analysis results show that YGN 3 and 4 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at worst time. Also, the YGN 3 and 4 RCP trip strategy demonstrates that both the 10 CFR 50.46 requirements on PCT and the ANSI standards 58.8 requirements on operator action time can be satisfied with enough margin. Therefore, it is concluded that the T2/L2 RCP trip strategy with a trip setpoint of 1775 psia for YGN 3 and 4 can provide improved operator guidance for the RCP operation during accidents. 11 figs., 4 tabs., 9 refs. (Author)