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Sample records for lpci

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

    International Nuclear Information System (INIS)

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

    2012-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  3. Evaluation report on CCTF Core-II reflood test C2-9 (Run 68)

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Iguchi, Tadashi; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Sugimoto, Jun.

    1987-02-01

    In order to study the LPCI flow rate effect on the core cooling and system behavior, a test was performed with the LPCI flow rate of 0.025 m 3 /s, which corresponds to the flow rate in case of no pump failure in a PWR system. Through the comparisons of test results with those from the reference test with the LPCI flow rate of 0.011 m 3 /s, the following conclusions were obtained: (1) The higher LPCI flow rate resulted in the worse core-cooling in these two tests. The test results show that the lower LPCI flow rate is not necessarily a conservative assumption for the evaluation of the core cooling during the reflood phase of a PWR LOCA. (2) The worse core-cooling in the high LPCI flow rate test is attributed to the lower core-pressure than in the reference test. It is found that the lower core-pressure results from the lower pressure drop through the broken cold leg. (3) It is expected that the current evaluation model(EM) code is still conservative because it usually predicts the low pressure drop through the broken cold leg. (4) The flow oscillation in the cold leg was not significant even in the high LPCI flow rate test before the whole core quench. (author)

  4. Using the Lunar Phases Concept Inventory to Investigate College Students' Pre-instructional Mental Models of Lunar Phases

    Science.gov (United States)

    Lindell, Rebecca S.; Sommer, Steven R.

    2004-09-01

    The Lunar Phases Concept Inventory (LPCI) is a twenty-item multiple-choice inventory developed to aid instructors in assessing the mental models their students utilize when answering questions concerning phases of the moon. Based upon an in-depth qualitative investigation of students' understanding of lunar phases, the LPCI was designed to take advantage of the innovative model analysis theory to probe the different dimensions of students' mental models of lunar phases. As part of a national field test, pre-instructional LPCI data was collected for over 750 students from multiple post-secondary institutions across the United States and Canada. Application of model analysis theory to this data set allowed researchers to probe the different mental models of lunar phases students across the country utilize prior to instruction. Results of this analysis display strikingly similar results for the different institutions, suggesting a potential underlying cognitive framework.

  5. Student Moon Observations and Spatial-Scientific Reasoning

    Science.gov (United States)

    Cole, Merryn; Wilhelm, Jennifer; Yang, Hongwei

    2015-07-01

    Relationships between sixth grade students' moon journaling and students' spatial-scientific reasoning after implementation of an Earth/Space unit were examined. Teachers used the project-based Realistic Explorations in Astronomical Learning curriculum. We used a regression model to analyze the relationship between the students' Lunar Phases Concept Inventory (LPCI) post-test score variables and several predictors, including moon journal score, number of moon journal entries, student gender, teacher experience, and pre-test score. The model shows that students who performed better on moon journals, both in terms of overall score and number of entries, tended to score higher on the LPCI. For every 1 point increase in the overall moon journal score, participants scored 0.18 points (out of 20) or nearly 1% point higher on the LPCI post-test when holding constant the effects of the other two predictors. Similarly, students who increased their scores by 1 point in the overall moon journal score scored approximately 1% higher in the Periodic Patterns (PP) and Geometric Spatial Visualization (GSV) domains of the LPCI. Also, student gender and teacher experience were shown to be significant predictors of post-GSV scores on the LPCI in addition to the pre-test scores, overall moon journal score, and number of entries that were also significant predictors on the LPCI overall score and the PP domain. This study is unique in the purposeful link created between student moon observations and spatial skills. The use of moon journals distinguishes this study further by fostering scientific observation along with skills from across science, technology, engineering, and mathematics disciplines.

  6. Evaluation report on CCTF Core-II reflood test C2-16 (Run 76)

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Hojo, Tsuneyuki; Murao, Yoshio; Sugimoto, Jun.

    1987-03-01

    This report presents the result of the upper plenum injection (UPI) test C2-16 (Run 76), which was conducted on October 23, 1984, with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute (JAERI). The CCTF is a 1/21.4 scale model of a 1,100 MWe PWR with four loop active components to provide information on the system and core thermo-hydrodynamics during reflood. The objectives of the test are to investigate the reflood phenomena with single failure UPI condition and to investigate the effect of the asymmetry of UPI on the reflood phenomena. The test was performed with an asymmetric UPI condition at the injection rate simulating single failure of LPCI pumps. It was observed that, (1) a UPI test simulating no LPCI pump failure gave the slightly lower peak clad temperature than a UPI test simulating single LPCI pump failure, indicating that single LPCI pump failure assumption is conserrative for UPI condition, and (2) an asymmetric UPI lead to a higher core water accumulation and then a higher heat transfer coefficient, resultantly a lower peak clad temperature than a symmetric UPI, indicating that asymmetric UPI does not lead to a poorer core cooling than symmetric UPI. (author)

  7. Assessment of core thermo-hydrodynamic models of REFLA-1D with CCTF data

    International Nuclear Information System (INIS)

    Okubo, Tsutomu; Murao, Yoshio

    1983-07-01

    In order to assess the core thermo-hydrodynamic models of REFLA-1D/MODE3, which is the latest version of REFLA-1D, several calculations of the core thermo-hydrodynamics have been performed for the CCTF Core-I series tests. The measured initial and boundary conditions were used for these calculations. The calculational results showed that the water accumulation model of Case 2 could predict the CCTF results fairly well as it could for the JAERI small scale facility. The calculated results for the base case and the EM tests were in good agreement with the CCTF data. The parameter effects, such as system pressure, initial clad temperature, Acc injection rate, LPCI injection rate and initial down-comer wall temperature, were predicted correctly, except for the high system pressure and the high LPCI injection rate tests. (author)

  8. Macroeconomic and Bank Specific Determinants of Non-Performing Loans (NPLs in the Indian Banking Sector

    Directory of Open Access Journals (Sweden)

    Memdani Laila

    2017-08-01

    Full Text Available The main objective of the paper is to find out the determinants of NPAs in the Indian Banking sector and to study if these determinants vary across the three different ownership structures viz., public sector banks (PSBs, private banks (PBs and foreign banks (FBs, of banks in India. The panel data for all the banks from 2005 to 2014 is collected from the official website of Reserve Bank of India (RBI, the Central Bank of the country. The econometric technique of Fixed Effects model and Random Effects model is used for the purpose. The results reveal that Macro economic factors, like log of percapita income (LPCY and Inflation (INFN, are significantly affecting NPLs in Public Sector Banks (PSBs. In case of private banks (PBs LPCY is highly significant while bank specific variables like size and total loans to total loans of the banking sector (TLTLBS are significant at 10% level. For FBs none of the variables were significant.

  9. Different immunomodulatory effects associated with sub-micrometer particles in ambient air from rural, urban and industrial areas

    International Nuclear Information System (INIS)

    Wichmann, Gunnar; Franck, Ulrich; Herbarth, Olf; Rehwagen, Martina; Dietz, Andreas; Massolo, Laura; Ronco, Alicia; Mueller, Andrea

    2009-01-01

    Immunomodulatory effects of chemicals adsorbed to particles with aerodynamic diameter below 0.49 μm (PM 0.5 ) collected in winter 2001 at three sampling points (industrial area [LPIn], traffic-influenced urban area [LPCi], and control area [LPCo]) of La Plata, Argentina, were investigated. The sampling of particulate matter was carried out with high-volume collectors using cascade impactors. PM 0.5 -adsorbed compounds were hexane-extracted by accelerated solvent extraction. For immunological investigations, human peripheral blood lymphocytes were activated by phytohemagglutinin and exposed to dimethyl-sulfoxide dilutions of PM 0.5 -extracts for 24 h. Vitality/proliferation was quantified using MTT, released interferon-γ (IFN-γ) and interleukin-4 (IL-4) by ELISA. Cytokine production but not vitality/proliferation was significantly suppressed by all of the highest extract concentrations. Generally, suppression of IFN-γ by PM 0.5 -extracts was stronger than those of IL-4. Based on administered mass of PM 0.5 , all extracts suppressed IFN-γ production nearly uniform. Contrary, LPCi-extracts exerted maximum IFN-γ suppression based either on air volume or regarding PM 0.5 -adsorbed PAH. Also the ranking of PM 0.5 -associated effects on IL-4 production differs in dependence of the chosen reference points, either mass or [μg/ml] or air volume [m 3 /ml] related dust quantities in cell culture. Based on the corresponding air volume, LPCi-extracts inhibited IL-4 production to the maximum extend, whereas suppression of IL-4 was comparable based on concentrations. This indicates that not only the disparate PM 0.5 -masses in air cause varying impacts, but also that disparities in PM 0.5 -adsorbed chemicals provoke different effects on immune responses and shifts in the regulatory balance that might have implications for allergy and cancer development

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

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

    1985-10-01

    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

  11. BWR 200 % recirculation pump suction line break LOCA tests, RUNs 942 and 943 at ROSA-III without HPCS

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Murata, Hideo; Koizumi, Yasuo

    1986-03-01

    This report presents the experimental results of RUNs 942 and 943 in ROSA-III program, which are 200 % recirculation pump suction line break LOCA tests with assumption of HPCS failure. The ROSA-III test facility simulates a BWR system with volume scale of 1/424 and has four half-length electrically heated fuel bundles, two active recirculation loops, ECCS's, and steam and feedwater systems. Effects of initial core void distribution and other fluid conditions on overall LOCA phenomena with special interest on transient core cooling phenomena were investigated by comparing the present test results with those of RUN 926, a 200 % suction line break test with standard initial fluid conditions. The initial core outlet quality was changed between 5 % and 43 %. As conclusions, (1) the initial lower core flow and higher void fraction affected significantly the core cooling conditions and resulted in earlier and higher PCT. (2) The lower plenum flashing temporarily contributed to cool down the core. (3) Flashing of remained hot water in the feedwater line affected slightly the pressure response and delayed the actuation of LPCI by 11 seconds. (4) The whole core was completely cooled down within 104 seconds after the LPCI actuation in these large break tests. (author)

  12. PRAAGE-1988: An interactive IBM-PC code for aging analysis of NUREG-1150 systems

    International Nuclear Information System (INIS)

    Fullwood, R.R.; Shier, W.G.

    1988-01-01

    Probabilistic Risk Assessments (PRA) contain a great deal of information for estimating the risk of a nuclear power plant but do not consider aging. PRAAGE (PRA+AGE) is an interactive, IBM-PC code for processing PRA-developed system models using non-aged failure rate data in conjunction with user-supplied time-dependent nuclear plant experience component failure rate data to determine the effects of component aging on a system's reliability as well as providing the age-dependent importances of various generic components. This paper describes the structure, use and application of PRAAGE to the aging analysis of the Peach Bottom 2 RHR system in the LPCI and SDC modes of operation. 4 refs., 15 figs., 5 tabs

  13. Experiment data of 200% recirculation pump discharge line break integral test run 961 with HPCS failure at ROSA-III and comparison with results of suction line break tests

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Nakamura, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Murata, Hideo; Yonomoto, Taisuke; Shiba, Masayoshi

    1984-03-01

    This report presents the experimental data of RUN 961, a 200% double-ended break test at the recirculation pump discharge line in the ROSA-III test facility. The ROSA-III test facility is a volumetrically scaled (1/424) system of the BWR/6. The facility has the electrically heated core, the break simulator and the scaled ECCS (Emergency Core Cooling System). The MSIV (Main Steam Isolation Valve) closure and the ECCS actuation were tripped by the liquid level in the upper downcomer. The PCT (Peak Cladding Temperature) was 894 K, which was 107 K higher than a 200% pump suction line break test (RUN 926) due to the smaller depressurization rate. The effect of break location on transient LOCA phenomena was clarified by comparing the discharge and suction break tests. The whole core was quenched 71 s after LPCI actuation and the effectiveness of ECCS has been confirmed. (author)

  14. Evaluation report on CCTF core-II reflood test C2 - 18 (Run 78)

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Murao, Yoshio; Sugimoto, Jun; Hojo, Tsuneyuki.

    1987-03-01

    This report presents the result of the upper plenum injection (UPI) test C2 - 18 (Run 78), which was conducted on November 13, 1984 with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute (JAERI). The CCTF is a 1/21.4 scale model of a 1,100 MWe PWR with four loop active components to provide information on the system and core thermo-hydrodynamics during reflood phase. The objectives of the test are to investigate the refill behavior with UPI condition and to investigate the reflood behavior with UPI Best-Estimate (BE) condition. The test was performed to simulate refill/reflood behavior with UPI and BE conditions (However, the LPCI flow rate was determined based on single failure of LPCI pumps.). The result of the test showed the followings. (1) Little special phenomena were recognized under UPI and BE conditions in comparison with those under UPI and Evaluation-Model (EM) conditions, although certain special phenoma (i.e. significant fluid oscillation) were recognized under Cold-Leg-Injection (CLI) and BE conditions in comparison with those under CLI and EM conditions. (2) Water inventory in lower plenum increased smoothly due to water injected into both upper plenum and cold leg during refill phase, similarly to that in refill-simulation test with CLI condition. Small difference in refill behavior with UPI condition is the existing of steam condensation in upper plenum, resulting in lower steam binding and higher core cooling during early reflood phase. This indicates the conservatism of UPI against CLI during early reflood phase. (3) The good core-cooling capability was confirmed under UPI and BE conditions. (author)

  15. Evaluating Middle School Students' Spatial-scientific Performance in Earth-space Science

    Science.gov (United States)

    Wilhelm, Jennifer; Jackson, C.; Toland, M. D.; Cole, M.; Wilhelm, R. J.

    2013-06-01

    Many astronomical concepts cannot be understood without a developed understanding of four spatial-mathematics domains defined as follows: a) Geometric Spatial Visualization (GSV) - Visualizing the geometric features of a system as it appears above, below, and within the system’s plane; b) Spatial Projection (SP) - Projecting to a different location and visualizing from that global perspective; c) Cardinal Directions (CD) - Distinguishing directions (N, S, E, W) in order to document an object’s vector position in space; and d) Periodic Patterns - (PP) Recognizing occurrences at regular intervals of time and/or space. For this study, differences were examined between groups of sixth grade students’ spatial-scientific development pre/post implementation of an Earth/Space unit. Treatment teachers employed a NASA-based curriculum (Realistic Explorations in Astronomical Learning), while control teachers implemented their regular Earth/Space units. A 2-level hierarchical linear model was used to evaluate student performance on the Lunar Phases Concept Inventory (LPCI) and four spatial-mathematics domains, while controlling for two variables (gender and ethnicity) at the student level and one variable (teaching experience) at the teacher level. Overall LPCI results show pre-test scores predicted post-test scores, boys performed better than girls, and Whites performed better than non-Whites. We also compared experimental and control groups’ by spatial-mathematics domain outcomes. For GSV, it was found that boys, in general, tended to have higher GSV post-scores. For domains CD and SP, no statistically significant differences were observed. PP results show Whites performed better than non-Whites. Also for PP, a significant cross-level interaction term (gender-treatment) was observed, which means differences in control and experimental groups are dependent on students’ gender. These findings can be interpreted as: (a) the experimental girls scored higher than the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    International Nuclear Information System (INIS)

    Ramirez G, C.; Chavez M, C.

    2012-10-01

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

  18. ROSA-III/971, BWR Rig of Safety Assessment LOCA, Loss of Offsite Power Transient

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: ROSA-III is a 1/124 scaled down test facility with electrically heated core designed to study the response of engineered safety features to loss-of-coolant accidents in in commercial BWR. It consists of the following, fully instrumented subsystems: (a) the pressure vessel with a core simulating four half-length fuel assemblies and control rod; (b) steam line and feed water line, which are independent open loops; (c) coolant recirculation system, which consists of two loops provided with a recirculation pump and two jet pumps in each loop; (d) emergency cooling system, including HPCS, LPCS, LPCI, and ADS. 2 - Description of test: Run 971 simulated a BWR LOSS of off-site power transient. The core scram was assumed to occur at 6 seconds after the transient initiated by the turbine trip. HPCS failure was assumed. After ADS started, the upper half of the core was uncovered by steam. The core was re-flooded by LPCS alone

  19. Accidents and transients analyses of a super fast reactor with single flow pass core

    International Nuclear Information System (INIS)

    Sutanto,; Oka, Yoshiaki

    2014-01-01

    Highlights: • Safety analysis of a Super FR with single flow pass core is conducted. • Loss of feed water flow leads to a direct effect on the loss of fuel channel flow. • The core pressure is sensitive to LOCA accidents due to the direct effect. • Small LOCA introduces a critical break. • The safety criteria for all selected events are satisfied. - Abstract: The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST

  20. Acoustic emission in wet or dry corrosion studies: an update on the art and usefulness of a data base; L'emission acoustique en corrosion humide et seche: etat de l'art et interet d'une base de donnees

    Energy Technology Data Exchange (ETDEWEB)

    Caron, D.; Rigault, C. [Cetim-Centre Technique des Industries Mecaniques, 30 - Senlis (France); Rothea, C.; Mazille, H. [Institut National des Sciences Appliquees, INSA, 69 - Villeurbanne (France); Gaillet, L.; Moulin, G.; Beranger, G. [Universite de Technologie de Compiegne, 60 (France)

    2001-07-01

    At the initiative of CETIM, a bibliographical work was accomplished at the LPCI (INSA Lyon) and UTC. The aim of this work was to establish the most current and complete state of the art on the potential of the acoustic emission (A.E.) for the detection of wet or dry corrosion met on many equipments used in the mechanical, chemical and petrochemical industries. Corrosion can affect all metallic equipments of industry under more or less aggressive operating conditions. In spite of the existing knowledge of the degradation mechanisms, the post-mortem analysis or laboratory simulation only gives either late or incomplete information. This usually results from difficulties to determine the actual local conditions. A.E. is a non-destructive control technique that can be applied as well in laboratory as on industrial site. According to literature data, the ability of A.E. to detect and monitor wet or dry corrosion was evidenced. Thus, it could be used for to a more reliable, secure and economic industrial management. However A.E. was and is still sometimes discussed. Indeed, the diversity of the equipments, processes and conditions of exploitation can generate some discrepancies. This study attempts to remove these ambiguities by detailing the principal parameters and key factors in order to objectively compare the referenced works. Our works confirms that A.E. is a very sensitive and powerful technique for studying wet or dry corrosion in laboratories or on industrial plants. (authors)

  1. Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

    International Nuclear Information System (INIS)

    Choi, Y.A.; Feltus, M.A.

    1995-01-01

    Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company's Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates

  2. Analysis of SCTF/CCTF counterpart test results

    International Nuclear Information System (INIS)

    Okubo, Tsutomu; Sobajima, Makoto; Iwamura, Takamichi; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi; Murao, Yoshio

    1990-06-01

    Slab Core Test Facility (SCTF) and Cylindrical Core Test Facility (CCTF) are large scale experimental facilities of Japan Atomic Energy Research Institute (JAERI) for the investigation of reflooding behavior during a postulated loss-of-coolant accident (LOCA) in PWRs. Although the flow area scaling ratios of both facilities to a 1,000 MWe class PWR are the same and 1/21.4, the SCTF has the same core width as the radius of the reference PWR while the CCTF has a 1/4.5 times shorter core radius. Therefore, a few SCTF/CCTF counterpart tests were conducted in order to investigate the difference in core reflooding behavior between in the SCTF and CCTF tests as well as the effect of core radial length on core two-dimensional thermo-hydrodynamic behavior. This report present the test results and an analysis on them. Major results obtained are: (1) Taking account of the differences in test conditions and facility design, core reflooding behavior is considered to be similar between the SCTF and the CCTF test. Main difference of the facility design is in the effective core flow area and this is considered to result in the difference in core water accumulation behavior. (2) The effect of core radial length on core two-dimensional thermo-hydrodynamic behavior has been observed to be significant and heat transfer enhancement or degradation in radial direction is more significant for the longer radius core. (3) In addition, where the core power varies significantly in the radial direction, significant heat transfer enhancement has been observed in the higher power bundle during the LPCI period. Also, in the peripheral region, heat transfer degradation has been observed more significantly in the outer bundle even they have the same bundle power. (4) Magnitude of these heat transfer enhancement or degradation was larger at the higher elevation than the midplane level in the SCTF test, whereas smaller in the CCTF test. (author)

  3. Evaluation report on SCTF Core-II test S2-08

    International Nuclear Information System (INIS)

    Ohnuki, Akira; Iwamura, Takamichi; Abe, Yutaka; Murao, Yoshio; Adachi, Hiromichi.

    1991-01-01

    The present report investigates the effects of the difference of the core inlet subcooling during reflood in a PWR-LOCA on the thermal-hydraulic behaviors including two-dimensional behaviors in the pressure vessel in the Slab Core Test Facility (SCTF) Core-II tests under gravity feed mode. The following test results are examined: Tests S2-02 (Reference test) and Test S2-08 (High subcooling test). The degree of the difference of the subcooling between the two tests was about 20 to 35 K in the LPCI period. The following conclusions were obtained from this study: (1) Higher the subcooling gave larger amount of water accumulation in the core and gave better core cooling. These tendencies were also recognized in comparisons under the same distance from the quench front. Since the same tendencies can be predicted in the analyses with REFLA code because of the lower steam generation rate below quench front in the high subcooling test, the differences in the tests are supposed to be caused by the same reason. (2) Higher the subcooling gave larger amount of water accumulation in upper plenum. The carry-over liquid mass into hot leg became smaller in the later period in the higher subcooling test. These differences for carry-over and de-entrainment characteristics can be explained by the differences of quench velocity and of steam mass flow rate generated in the core. (3) No significant influence of the different degree of the subcooling was observed on the two-dimensional thermal-hydraulic behaviors in the pressure vessel. Namely, radial differences of sectional void fraction, heat transfer coefficient and the pressure among bundles at the same elevation were almost the same amount for the two tests. Radial differences of liquid levels in the upper plenum was also almost the same amount for the two tests. (J.P.N.)

  4. Evaluation report on SCTF Core-II test S2-19

    International Nuclear Information System (INIS)

    Ohnuki, Akira; Iwamura, Takamichi; Iguchi, Tadashi; Abe, Yutaka; Murao, Yoshio; Adachi, Hiromichi.

    1991-03-01

    Experimental studies using Slab Core Test Facility (SCTF) have revealed that the heat transfer enhancement in higher power bundles is mainly governed by the radial power ratio in core during the reflood in PWR-LOCA. As a physical mechanism for the heat transfer enhancement, it can be considered from the experimental evidence that the increase of upward steam flow rate in a higher power bundle which is caused by the higher steam production rate in the bundle gives the higher upward liquid flow rate in the bundle and the increase of the liquid flow rate gives the heat transfer enhancement. In order to develop a mechanistic model for the heat transfer enhancement based on this idea, the following relations should be identified quantitatively: (1) Relation between the steam production rate and the upward liquid flow rate, (2) Cross flow rate above the quench front and (3) Relation between the degree of heat transfer enhancement due to radial power ratio and the amount of increase of upward liquid flow rate. In this report, the above relation (3) was investigated experimentally as a step to develop the mechanistic model using the SCTF where the relation between the radial power ratio and the heat transfer enhancement has been made clear quantitatively. The degree of increase of heat transfer between two forced feed tests with the different flow rate in LPCI period was compared with the degree of heat transfer enhancement under a radial power ratio in the previous SCTF tests. The two forced feed tests were performed under the condition without any significant two-dimensional hydraulic behavior in core. The ratio of the mass flow rate between the two tests was about double. (author)

  5. Reflood behavior at low initial clad temperature in Slab Core Test Facility Core-II

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Sobajima, Makoto; Abe, Yutaka; Iwamura, Takamichi; Ohnuki, Akira; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Adachi, Hiromichi.

    1990-07-01

    In order to study the reflood behavior with low initial clad temperature, a reflood test was performed using the Slab Core Test Facility (SCTF) with initial clad temperature of 573 K. The test conditions of the test are identical with those of SCTF base case test S2-SH1 (initial clad temperature 1073 K) except the initial clad temperature. Through the comparison of results from these two tests, the following conclusions were obtained. (1) The low initial clad temperature resulted in the low differential pressures through the primary loops due to smaller steam generation in the core. (2) The low initial clad temperature caused the accumulated mass in the core to be increased and the accumulated mass in the downcomer to be decreased in the period of the lower plenum injection with accumulator (before 50s). In the later period of the cold leg injection with LPCI (after 100s), the water accumulation rates in the core and the downcomer were almost the same between both tests. (3) The low initial clad temperature resulted in the increase of the core inlet mass flow rate in the lower plenum injection period. However, the core inlet mass flow rate was almost the same regardless of the initial clad temperature in the later period of the cold leg injection period. (4) The low initial clad temperature resulted in the low turnaround temperature, high temperature rise and fast bottom quench front propagation. (5) In the region apart from the quench front, low initial clad temperature resulted in the lower heat transfer. In the region near the quench front, almost the same heat transfer coefficient was observed between both tests. (6) No flow oscillation with a long period was observed in the SCTF test with low initial clad temperature of 573 K, while it was remarkable in the Cylindrical Core Test Facility (CCTF) test which was performed with the same initial clad temperature. (J.P.N.)

  6. Use and Evaluation of 3D GeoWall Visualizations in Undergraduate Space Science Classes

    Science.gov (United States)

    Turner, N. E.; Hamed, K. M.; Lopez, R. E.; Mitchell, E. J.; Gray, C. L.; Corralez, D. S.; Robinson, C. A.; Soderlund, K. M.

    2005-12-01

    One persistent difficulty many astronomy students face is the lack of 3- dimensional mental model of the systems being studied, in particular the Sun-Earth-Moon system. Students without such a mental model can have a very hard time conceptualizing the geometric relationships that cause, for example, the cycle of lunar phases or the pattern of seasons. The GeoWall is a recently developed and affordable projection mechanism for three-dimensional stereo visualization which is becoming a popular tool in classrooms and research labs for use in geology classes, but as yet very little work has been done involving the GeoWall for astronomy classes. We present results from a large study involving over 1000 students of varied backgrounds: some students were tested at the University of Texas at El Paso, a large public university on the US-Mexico border and other students were from the Florida Institute of Technology, a small, private, technical school in Melbourne Florida. We wrote a lecture tutorial-style lab to go along with a GeoWall 3D visual of the Earth-Moon system and tested the students before and after with several diagnostics. Students were given pre and post tests using the Lunar Phase Concept Inventory (LPCI) as well as a separate evaluation written specifically for this project. We found the lab useful for both populations of students, but not equally effective for all. We discuss reactions from the students and their improvement, as well as whether the students are able to correctly assess the usefullness of the project for their own learning.

  7. Safety analysis of high temperature reactor cooled and moderated by supercritical light water

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki; Oka, Yoshiaki; Koshizuka, Seiichi

    2003-01-01

    This paper describes 'Safety' of a high temperature supercritical light water cooled and moderated reactor (SCRLWR-H) with descending flow water rods. The safety system of the SCLWR-H is similar to that of a BWR. It consists of reactor scram, high pressure auxiliary feedwater system (AFS), low pressure core injection system (LPCI), safety relief valves (SRV), automatic depressurization system (ADS), and main steam isolation valves (MSIV). Ten types of transients and five types of accidents are analyzed using a plant transient analysis code SPRAT-DOWN. The sequences are determined referring to LWRs. At the 'Loss of load without turbine bypass' transient, the coolant density and the core power are increased by the over-pressurization, and at the same time the core flow rate is decreased by the closure of the turbine control valves. The peak cladding temperature increases to 727degC. The high temperature at this type of transient is one of the characteristics of the SCLWR-H. Conversely at 'feedwater-loss' events, the core power decrease to some extend by density feedback before the reactor scram. The peak cladding temperatures at the 'Partial loss of feedwater' transient and the 'Total loss of feedwater' accident are only 702degC and 833degC, respectively. The cladding temperature does not increase so much at the transients 'Loss of feedwater heating' and 'CR withdrawal' because of the operation of the plant control system. All the transients and accidents satisfy the satisfy criteria with good margins. The highest cladding temperatures of the transients and the accidents are 727degC and 833degC at the 'Loss of load without turbine bypass' and 'Total loss of feedwater', respectively. The duration of the high cladding temperature is very short at the transients. According to the parametric survey, the peak cladding temperature are sensitive to the parameters such as the pump coast-down time, delay of pump trip, AFS capacity, AFS delay, CR worth, and SRV setpoint