Experimental study on reflooding in advanced tight lattice PWR

International Nuclear Information System (INIS)

Hori, K.; Kodama, J.; Teramae, T.

2000-01-01

This paper is related to the experimental study on the feasibility of core cooling by re-flooding in a large break loss of coolant accident (LOCA) for the advanced tight lattice pressurized water reactor (PWR). The tight lattice core design should be adopted to improve the conversion ratio. Major one of the key questions of such tight lattice core is the cooling capability under the re-flood condition in a large break LOCA. Forced feed bottom re-flooding experiments have been performed by use of a 4x4 triangular array rod bundle. The rod gap is 0.5 mm, 1.0 mm, or 1.5 mm. The measured peak temperature is below around 1273 K even in case of 1.0/0.5 mm rod gap. And, the evaluation based on the experimental results of rod temperatures and core pressure drop also shows that the core cooling under re-flooding condition is feasible. (author)

Assessment of one dimensional reflood model in REFLA/TRAC code

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

1993-12-01

Post-test calculations for twelve selected SSRTF, SCTF and CCTF tests were performed to assess the predictive capability of the one-dimensional reflood model in the REFLA/TRAC code for core thermal behavior during the reflood in a PWR LOCA. Both core void fraction profile and clad temperature transients were predicted excellently by the REFLA/TRAC code including parameter effect of core inlet subcooling, core flooding rate, core configuration, core power, system pressure, initial clad temperature and so on. The peak clad temperature was predicted within an error of 50 K. Based on these assessment results, it is verified that the core thermal hydraulic behaviors during the reflood can be predicted excellently with the REFLA/TRAC code under various conditions where the reflood may occur in a PWR LOCA. (author)

2012 SARA Students Technical Report

International Nuclear Information System (INIS)

Briccetti, Angelo; Lorei, Nathan; Yonkings, David; Lorio, David; Goorley, John T.; Sood, Avneet

2012-01-01

The Service Academy Research Associates (SARA) program provides an opportunity for Midshipmen and Cadets from US Service Academies to participate in research at Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratory for several weeks during the summer as part of their summer training assignments. During the summer of 2012, three Midshipmen were assigned to work with the XCP Division at LANL for approximately 5-6 weeks. As one of the nation's top national security science laboratories, LANL stretches across 36 square miles, has over 2,100 facilities, and employs over 9,000 individuals including a significant number of students and postdocs. LANL's mission is to 'apply science and technology to: ensure the safety, security, and reliability of the US nuclear deterrent, reduce global threats, and solve other emerging national security challenges.' While LANL officially operates under the US Department of Energy (DoE), fulfilling this mission requires mutual cooperation with the US Department of Defense (DoD) as well. LANL's high concentration of knowledge and experience provides interns a chance to perform research in many disciplines, and its connection with the DoD in both operation and personnel gives SARA students insight to career possibilities both during and after military service. SARA students have plenty of opportunity to enjoy hiking, camping, the Los Alamos YMCA, and many other outdoor activities in New Mexico while staying at the Buffalo Thunder Resort, located 20 miles east of the lab. XCP Division is the Computational Physics division of LANL's Weapons Department. Working with XCP Division requires individuals to be Q cleared by the DoE. This means it is significantly more convenient for SARA students to be assigned to XCP Division than their civilian counterparts as the DoD CNWDI clearance held by SARA students is easily transferred to the lab prior to the students arriving at the start of

2012 SARA Students Technical Report

Energy Technology Data Exchange (ETDEWEB)

Briccetti, Angelo [Los Alamos National Laboratory; Lorei, Nathan [Los Alamos National Laboratory; Yonkings, David [Los Alamos National Laboratory; Lorio, David [Los Alamos National Laboratory; Goorley, John T. [Los Alamos National Laboratory; Sood, Avneet [Los Alamos National Laboratory

2012-07-30

The Service Academy Research Associates (SARA) program provides an opportunity for Midshipmen and Cadets from US Service Academies to participate in research at Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratory for several weeks during the summer as part of their summer training assignments. During the summer of 2012, three Midshipmen were assigned to work with the XCP Division at LANL for approximately 5-6 weeks. As one of the nation's top national security science laboratories, LANL stretches across 36 square miles, has over 2,100 facilities, and employs over 9,000 individuals including a significant number of students and postdocs. LANL's mission is to 'apply science and technology to: ensure the safety, security, and reliability of the US nuclear deterrent, reduce global threats, and solve other emerging national security challenges.' While LANL officially operates under the US Department of Energy (DoE), fulfilling this mission requires mutual cooperation with the US Department of Defense (DoD) as well. LANL's high concentration of knowledge and experience provides interns a chance to perform research in many disciplines, and its connection with the DoD in both operation and personnel gives SARA students insight to career possibilities both during and after military service. SARA students have plenty of opportunity to enjoy hiking, camping, the Los Alamos YMCA, and many other outdoor activities in New Mexico while staying at the Buffalo Thunder Resort, located 20 miles east of the lab. XCP Division is the Computational Physics division of LANL's Weapons Department. Working with XCP Division requires individuals to be Q cleared by the DoE. This means it is significantly more convenient for SARA students to be assigned to XCP Division than their civilian counterparts as the DoD CNWDI clearance held by SARA students is easily transferred to the lab prior to the

Large scale reflood test

International Nuclear Information System (INIS)

Hirano, Kemmei; Murao, Yoshio

1980-01-01

The large-scale reflood test with a view to ensuring the safety of light water reactors was started in fiscal 1976 based on the special account act for power source development promotion measures by the entrustment from the Science and Technology Agency. Thereafter, to establish the safety of PWRs in loss-of-coolant accidents by joint international efforts, the Japan-West Germany-U.S. research cooperation program was started in April, 1980. Thereupon, the large-scale reflood test is now included in this program. It consists of two tests using a cylindrical core testing apparatus for examining the overall system effect and a plate core testing apparatus for testing individual effects. Each apparatus is composed of the mock-ups of pressure vessel, primary loop, containment vessel and ECCS. The testing method, the test results and the research cooperation program are described. (J.P.N.)

Reflooding phase of the LOCA - state of the art I. Heat transfer and fluid flow during reflooding

International Nuclear Information System (INIS)

Yadigaroglu, G.

1977-01-01

Complex heat transfer processes take place during the reflooding phase of the Loss-of-Coolant Accident in Light-Water Reactors. Reflooding experiments conducted with simple single-channel geometries (round tubes and annuli) and with rod bundles are reviewed. The experimental findings and various parametric trends are critically discussed, explained, and summarized. Analytical methods that are in use in safety analysis and features of advanced models that have been proposed are outlined. These advanced models attempt to solve the conservation equations in the core channels in order to find the local coolant conditions. The values of the heat transfer coefficients are related to local parameters to arrive at cladding temperature predictions

RELAP/REFLA (Mod 0): a system reflooding analysis computer program

International Nuclear Information System (INIS)

Fujiki, Kazuo; Murao, Yoshio; Shimooke, Takanori.

1981-03-01

A new computer code RELAP/REFLA has been developed, aiming at analyses of the core reflooding phenomena during the postulated loss-of-coolant accident of PWRs. The code was originated from the combination of two distinct codes, RELAP4-FLOOD and REFLA-1D. The characteristics of the code are: (1) Kinematical model based on the observation and analysis of quench experiments is used for the thermal-hydraulic analysis of reflooding core, (2) it has the capability to analyse the reflooding phenomena in an arbitrary type of PWR or experimental facility, including the system feedback effects, (3) the flow paths in the actual system are represented by the combination of 1-dimentional flow paths, and vapor-liquid equilibrium model is applied except the reflooding core. This report is a code manual of RELAP/REFLA (version Mod 0) and contains the descriptions of the basic models, basic equations, code structure and input format. The calculated results of two kinds of sample problems, i.e., reflooding problem on the 4 loop PWR and FLECHT-SET experiment, are also presented. Relatively close agreement between FLECHT-SET data and the calculated results was obtained for the lower portion of the core, but poor agreement for the temperature histories in the upper core and carryover ratio. Running speed and core memory size are almost equal to those of RELAP 4/Mod 3. (author)

Report on series 2B reflood experiment

International Nuclear Information System (INIS)

Murao, Yoshio; Iguchi, Tadashi; Sudoh, Takashi; Sudo, Yukio; Sugimoto, Jun

1976-12-01

Series 2B reflood experiment was carried out from April to May 1975, as follows: 1) injection of coolant water from the downcomer at a constant head into the test section having a flow resistance simulator of the primary loop, 2) under an atmospheric pressure, 3) in constant power density, 4) with heater rod temperature up to 600 0 C. The objectives are to examine quantitatively system effect and to check performance of the reflood test rig. The effect of the coolant injection mode, relation between oscillatory phenomena and core thermo-hydrodynamics, and technological problems of the test rig were observed. (auth.)

The SARA Consortium: Providing Undergraduate Access to a 0.9-m Telescope at Kitt Peak National Observatory

Science.gov (United States)

Wood, M. A.

2003-12-01

The Southeastern Research for Astronomy (SARA) operates a 0.9-m telescope at Kitt Peak National Observatory (KPNO). The member institutions are Florida Institute of Technology, East Tennessee State University, Florida International University, The University of Georgia at Athens, Valdosta State University, and Clemson University. The NSF awarded the KPNO #1 0.9-m telescope to the SARA Consortium in 1990. We built a new facility and began routine on-site observations in 1995. We began routine remote observations in 1999 using VNC to export the telescope and CCD control screens, and a web-cam in the dome to provide critical visual feedback on the status of the telescope and dome. The mission of the SARA Consortium is to foster astronomical research and education in the Southeastern United States. Although only two of the member institutions have no graduate programs, all six have a strong emphasis on undergraduate research and education. By pooling our resources, we are able to operate a research-grade facility that none of the individual schools could manage by itself, and in the process we can offer our undergraduate students the opportunity to assist in our research projects as well as to complete their own independent research projects using a facility at a premier site. The SARA Consortium also hosts a NSF REU Summer Intern Program in Astronomy, in which we support 11-12 students that work one-on-one with a SARA faculty mentor. Most of these interns are selected from primarily undergraduate institutions, and have not had significant previous research experience. As part of the program, interns and mentors travel to KPNO for a 4-5 night observing run at the telescope. The SARA NSF REU Program is funded through NSF grant AST-0097616.

Reactor hydrodynamics during the reflood phase of a loss-of-coolant accident

International Nuclear Information System (INIS)

Gay, R.R.

1977-01-01

The thermohydraulics of a nuclear reactor during the reflood phase of a hypothetical loss-of-coolant accident can be represented by moving control volume methodology in which six control volumes are used to represent the downcomer, lower plenum, and reactor core. The one-dimensional, homogeneous, equilibrium constitutive equations for two-phase steam/water flow are solved in each control volume and connecting junctions. One of the three core control volumes represents the quench region; it changes size and position based on the axial location of the clad quench temperature and the condensed liquid level in the flow channel. The lengths of the remaining two core control volumes are determined by the position of the quench region. Simulation of actual reflood experiments demonstrates that the methodology predicts reflood-like flow oscillations and reproduces the correct trends in experimental data. The moving control volume methodology has proven itself as a valid concept for reflood hydrodynamics, but further development of the existing EFLOD code is required for simulation of actual reflood experiments

Severe accident recriticality analyses (SARA)

DEFF Research Database (Denmark)

Frid, W.; Højerup, C.F.; Lindholm, I.

2001-01-01

with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality-both super-prompt power bursts and quasi steady-state power......Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies......, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g(-1), was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s(-1). In most cases, however, the predicted energy deposition was smaller, below...

Reflooding experiments on a 49-rod cluster containing a long 90% blockage

International Nuclear Information System (INIS)

Pearson, K.G.; Cooper, C.A.; Jowitt, D.; Kinneir, J.H.

1983-01-01

A series of reflooding experiments was performed on a model fuel assembly, containing a very severe partial blockage, in the THETIS rig. The assembly comprised 49 full length, electrically heated fuel rod simulators and the blockage was created by attaching thin-walled, preformed swellings to a group of 16 rods. Results are presented for single phase and forced reflooding experiments. The most important findings relate to the improvements in heat transfer created by spacer grids and the nature of the heat transfer processes within the blockage. Spacer grids are shown to improve heat transfer by increasing turbulence and also, when wet, by cooling the steam flowing through them. Liquid penetration evidently deteriorates as the rewetting front approaches the blockage, allowing the steam through the blockage to superheat strongly and giving rise to a late peak in cladding temperature. At low reflooding rates there is a temperature penalty associated with the blockage which becomes increasingly larger as the reflooding rate is reduced. The adequacy of cooling in this very severe blockage becomes questionable when the reflooding rate falls to about 2cm/s. (U.K.)

Improvements to the RELAP5/MOD3 reflood model and uncertainty quantification of reflood peak clad temperature

International Nuclear Information System (INIS)

Lee, Sang Yong; Chung, Bob Dong; Lee, Young Jin; Park, Chan Eok; Lee, Guy Hyung; Choi, Chul Jin

1994-06-01

This research aims to develop reliable, advanced system thermal-hydraulic computer code and to quantify the uncertainties of code to introduce the best estimate methodology of ECCS for LBLOCA. Although the one of best estimate code, RELAP5/MOD3.1 was introduced from USNRC, several deficiencies in its reflood model and some improvements have been made. The improvements consist of modification of reflood wall heat transfer package and adjusting the drop size in dispersed flow regime. The tome smoothing of wall vaporization and level tracking model are also added to eliminate the pressure spike and level oscillation. For the verification of improved model and quantification of associated uncertainty, the FLECHT-SEASET data were used and upper limit of uncertainty at 95% confidence level is evaluated. (Author) 30 refs., 49 figs., 2 tabs

Improvements to the RELAP5/MOD3 reflood model and uncertainty quantification of reflood peak clad temperature

Energy Technology Data Exchange (ETDEWEB)

Lee, Sang Yong; Chung, Bob Dong; Lee, Young Jin; Park, Chan Eok; Lee, Guy Hyung; Choi, Chul Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

This research aims to develop reliable, advanced system thermal-hydraulic computer code and to quantify the uncertainties of code to introduce the best estimate methodology of ECCS for LBLOCA. Although the one of best estimate code, RELAP5/MOD3.1 was introduced from USNRC, several deficiencies in its reflood model and some improvements have been made. The improvements consist of modification of reflood wall heat transfer package and adjusting the drop size in dispersed flow regime. The tome smoothing of wall vaporization and level tracking model are also added to eliminate the pressure spike and level oscillation. For the verification of improved model and quantification of associated uncertainty, the FLECHT-SEASET data were used and upper limit of uncertainty at 95% confidence level is evaluated. (Author) 30 refs., 49 figs., 2 tabs.

Assessment of reflood heat transfer model of COBRA-TIF against ABB-CE evaluation model

Energy Technology Data Exchange (ETDEWEB)

Lee, S. I.; Lee, S. Y.; Park, C. E.; Choi, H. R.; Choi, C. J. [Korea Power Engineering Company Inc., Taejon (Korea, Republic of)

2000-05-01

According to 10 CFR 50 Appendix K, ECCS performance evaluation model should be based on the experimental data of FLECHT and have the conservatism compared with experimental data. To meet this requirement ABB-CE has the complicate code structure as follows: COMPERC-II calculates three reflood rates, and FLELAPC and HTCOF calculate the reflood heat transfer coefficients, and finally STRIKIN-II calculates the cladding temperature using the reflood heat transfer calculated in previous stage. In this paper, to investigate whether or not COBRA-TF satisfies the requirement of Appendix K, the reflood heat transfer coefficient of COBRA-TF was assessed against ABB-CE MOD-2C model. It was found out that COBRA-TF predicts properly the experimental data and has more conservatism than the results of ABB-CE MOD-2C model. Based on these results, it can be concluded that the reflood heat transfer coefficients calculated by COBRA-TF meet the requirement of Appendix K.

News developments at Sara

International Nuclear Information System (INIS)

Belmont, J.L.; Fruneau, M.; Martin, P.

1989-05-01

SARA was one of the first cyclotrons to operate with an ECR ion source. Experience since 1983 showed it would be useful to have two sources situated outside the cyclotron vault to manage continuous operation of the accelerator and development of new ions. A 18 m long injection line for FERROMAFIOS and MINIMAFIOS has been constructed. One of its principal features is a 11 m long electrostatic guide with periodic focusing. Transmission ratios from the source to the internal beam are close to 30%. Experiences involving time of flight measurements require short beam bunches; in order to reduce their duration, a phase selecting system will be installed in the centre of the injector of SARA. A RF voltage at four times the frequency of the dees will vertically deflect early orbits while particles close to the central phase will cross the deflector at zero voltage and will undergo normal acceleration

Interpreting the SARA and RCRA training requirements

International Nuclear Information System (INIS)

Moreland, W.M.; Wells, S.M.

1987-01-01

The Resource Conservation and Recovery Act (RCRA) and the Superfund Amendments and Reauthorization Act (SARA) promulgated by the EPA (RCRA) and the OSHA (SARA) require hazardous materials training for all individuals working with hazardous materials. Facilities that are involved in the generation, storage, treatment, transportation, or disposal/removal of hazardous materials/waste must comply with all relevant training regulations. Using the guidelines contained in the RCRA and SARA regulations, decisions must be made to determine: the type of regulatory requirement based on facility function (i.e., whether the facility is a RCRA or CERCLA facility). The type of training required for specific categories of workers (e.g. managers, supervisors, or general site workers). The level of training needed for each category of worker. This presentation outlines how the Environmental Compliance and Health Protection Technical Resources and Training Group, working with waste operations personnel, establishes specific training requirements

Subacute ruminal acidosis (SARA) in grazing Irish dairy cows.

Science.gov (United States)

O'Grady, Luke; Doherty, Michael L; Mulligan, Finbar J

2008-04-01

Subacute ruminal acidosis (SARA) is a significant production disease of dairy cattle. Previous concerns have been raised over the occurrence of SARA in pasture-fed dairy cattle and the potential consequences of laminitis and lameness. Highly digestible perennial rye grass contains high concentrations of rapidly fermentable carbohydrate and low concentrations of physical effective fibre that may result in SARA. This study conducted a point prevalence survey of rumen health status in grazing Irish dairy cattle fed predominantly perennial rye grass-based pasture. The survey assessed rumen fluid, animal health status, milk production data and pasture composition. A total of 144 cows between 80 and 150 days in milk were sampled on 12 farms. Eleven percent of cows were classified as affected with SARA (pH 5.8). The study showed that low rumen pH is prevalent in grazing Irish dairy cattle consuming perennial rye grass-based pasture and raises concerns regarding effective pasture utilisation and possible consequences for animal health.

UC-B reflood experimental plan

International Nuclear Information System (INIS)

Yu, K.P.; Abdollahian, D.; Peake, W.T.; Elias, E.; Yadigaroglu, G.; Greif, R.

1977-04-01

The EPRI sponsored single tube reflooding heat transfer facility is described. The facility is located at the University of California, Berkeley. The physical systems which constitute the facility as well as the objectives and background of the program are described. The steam-water separator is described in detail Finally, the operating procedure and the test apparatus performance are discussed

System Analysis and Risk Assessment system (SARA) Version 4.0

International Nuclear Information System (INIS)

Sattison, M.B.; Russell, K.D.; Skinner, N.L.

1992-01-01

This NUREG is the tutorial for the System Analysis and Risk Assessment System (SARA) Version 4.0, a microcomputer-based system used to analyze the safety issues of a family [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. A series of lessons are provided that walk the user through some basic steps common to most analyses performed with SARA. The example problems presented in the lessons build on one another, and in combination, lead the user through all aspects of SARA sensitivity analysis

The effect of inlet flow oscillations on reflooding of a tubular test section

International Nuclear Information System (INIS)

Oh, S.; Banerjee, S.; Yadigaroglu, G.

1983-01-01

The reflooding of a vertical channel under oscillatory inlet flow conditions has been investigated experimentally. Compared to constant injection, oscillations always increase the liquid carryover in the early stages of reflooding. As reflooding progresses, the enhancement diminishes. The crossover point roughly coincides with saturation of the liquid at the quench front (QF). The higher liquid carryover at the beginning increases downstream heat transfer and speeds up QF propagation. But this higher liquid carryover, in turn, reduces the test section mass accumulation rate and delays QF propagation at later stages. The enhancement of liquid carryover, and the early increase and subsequent decrease in quench velocity are all accentuated as the oscillation amplitude and frequency increase. Large amplitude oscillations change the characteristics of QF propagation and the heat transfer immediately downstream of QF substantially. Correlations based on constant-injection reflooding data are not adequate, even if they are applied on an average local-conditions basis

Data report on series 6 reflood experiment

International Nuclear Information System (INIS)

Iguchi, Tadashi; Murao, Yoshio; Sudoh, Takashi; Sudo, Yukio; Sugimoto, Jun

1979-03-01

Series 6 reflood experiments (experiment with 4 x 4 indirectly heated rods) were carried out from March to June 1978. The purpose of the experiments was: 1) to observe overall reflood phenomena in a 4 x 4 indirectly heated heater rod bundle with thermocouples inbedded completely in the cladding, 2) to examine the quench characteristics at low flooding rate, 3) to measure steady-state differential pressures in the core, 4) to investigate the heat transfer coefficients before quenching, 5) to investigate the water effluence behavior at outlet of the core, 6) to investigate the effect of a non-heated rod, and 7) to examine the response characteristics of the system at forced oscillating flooding rate. Described are the experimental conditions and the results (cladding temperatures, pressure differences and flow rates) in the constant flooding rate experiments in series 6 experiment. (author)

Report on series 2A reflood experiment

International Nuclear Information System (INIS)

Murao, Yoshio; Iguchi, Tadashi; Sudoh, Takashi; Sudo, Yukio; Sugimoto, Jun

1976-11-01

Series 2A reflood experiment was carried out from February to April 1975 to obtain thermo-hydrodynamic data during reflood phase of a typical PWR. The main test conditions are as follows: - direct water injection into the simulated core at constant flow rate - operation under an atmospheric pressure, and - temperature of heater rods up to 600 0 C. Study of the data showed that several heat transfer phases exist in the core, i.e. adiabatic, droplet-dispersed vapor flow, film boiling, quench, and nucleate boiling phase. The relation between heat transfer phases and heat transfer coefficients was discussed qualitatively, and the following phenomena were found out: Pressure oscillation exists in the core, and it has large influence upon heat transfer coeficient characteristic as well as heater rod surface temperature response, and the inlet water velocity influences the carry over fraction. (auth.)

The relationship between SARA fractions and crude oil stability

Directory of Open Access Journals (Sweden)

Siavash Ashoori

2017-03-01

Full Text Available Asphaltene precipitation and deposition are drastic issues in the petroleum industry. Monitoring the asphaltene stability in crude oil is still a serious problem and has been subject of many studies. To investigate crude oil stability by saturate, aromatic, resin and asphaltene (SARA analysis seven types of crudes with different components were used. The applied methods for SARA quantification are IP-143 and ASTM D893-69 and the colloidal instability index (CII is computed from the SARA values as well. In comparison between CII results, the values of oil compositions demonstrated that the stability of asphaltenes in crude oils is a phenomenon that is related to all these components and it cannot be associated only with one of them, individually.

Improved guidelines for RELAP4/MOD6 reflood calculations

International Nuclear Information System (INIS)

Chen, T.H.; Fletcher, C.D.

1980-01-01

Computer simulations were performed for an extensive selection of forced- and gravity-feed reflood experiments. This effort was a portion of the assessment procedure for the RELAP4/MOD6 thermal hydraulic computer code. A common set of guidelines, based on recommendations from the code developers, was used in determining the model and user-selected input options for each calculation. The comparison of code-calculated and experimental data was then used to assess the capability of the RELAP4/MOD6 code to model the reflood phenomena. As a result of the assessment, the guidelines for determining the user-selected input options were improved

Phosphorus Dynamics in Long-Term Flooded, Drained, and Reflooded Soils

Directory of Open Access Journals (Sweden)

Juan Tian

2017-07-01

Full Text Available In flooded areas, soils are often exposed to standing water and subsequent drainage, thus over fertilization can release excess phosphorus (P into surface water and groundwater. To investigate P release and transformation processes in flooded alkaline soils, wheat-growing soil and vegetable-growing soil were selected. We flooded-drained-reflooded two soils for 35 d, then drained the soils, and 10 d later reflooded the soils for 17 d. Dissolved reactive phosphorus (DRP, soil inorganic P fractions, Olsen P, pH, and Eh in floodwater and pore water were analyzed. The wheat-growing soil had significantly higher floodwater DRP concentrations than vegetable-growing soil, and floodwater DRP in both soils decreased with the number of flooding days. During the reflooding period, DRP in overlying floodwater from both soils was less than 0.87 mg/L, which was 3–25 times less than that during the flooding period. Regardless of flooding or reflooding, pore water DRP decreased with flooding days. The highest concentration of pore water DRP observed at a 5-cm depth. Under the effect of fertilizing and flooding, the risk of vertical P movement in 10–50 cm was enhanced. P diffusion occurred from the top to the bottom of the soils. After flooding, Al-P increased in both soils, and Fe-P, O-P, Ca2-P decreased, while Fe-P, Al-P, and O-P increased after reflooding, When Olsen P in the vegetable-growing soil exceeded 180.7 mg/kg and Olsen P in the wheat-growing soil exceeded 40.8 mg/kg, the concentration of DRP in pore water increased significantly. Our results showed that changes in floodwater and pore water DRP concentrations, soil inorganic P fractions, and Olsen P are significantly affected by fertilizing and flooding; therefore, careful fertilizer management should be employed on flooded soils to avoid excess P loss.

Evaluation of the pressure difference across the core during PWR-LOCA reflood phase

International Nuclear Information System (INIS)

Iguchi, Tadashi; Murao, Yoshio

1979-03-01

The flooding rate of the core influences largely cooling of the core during the reflood phase of a PWR-LOCA. Since the void fraction of two-phase flow in the core is important determining the flooding rate, it is essential to examine this void fraction. The void fraction in the core during the reflood phase obtained by experiment was compared with those predicted by the correlations respectively of Akagawa, Nicklin, Zuber, Yeh, Griffice, Behringer and Jhonson. Only Yeh's correlation was found to be usable for the purpose. The pressure difference of the core during the reflood phase was calculated by reflood analyzing code REFLA-1D using Yeh's correlation. Following are the results: (1) During the steady-state period after quenching of the heaters, the prediction agrees within +-15% with the experiment. (2) During the transient period when the quench front is advancing, the prediction is not in agreement with the experiment, the difference being about +-40%. Influence of the advancing quench front upon the void fraction in the core must further be studied. (author)

RELAP4/MOD6 analysis of forced- and gravity-feed reflood tests

International Nuclear Information System (INIS)

Chen, T.H.; Fletcher, C.D.

1980-01-01

The RELAP4/MOD6 computer code is used for the analysis of the reactor core heat transfer during the reflooding phase of a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). The code requires the user to specify input parameters for the reflood heat transfer models. Results of previous comparisons of code calculations with experimental data have indicated no single selection of input parameters is adequate for a spectrum of tests and test facilities. This paper presents the development of revised quidelines and assesses the effect of those modifications on RELAP4/MOD6 data comparisons using previously analyzed reflood experiments. The paper also presents an assessment of the revised guidelines and the original guidelines against experimental data significantly different from previously analyzed tests

Layering of life (Sara novel of Peter Sarić

Directory of Open Access Journals (Sweden)

Kostić Dragomir J.

2015-01-01

Full Text Available Novel Sara of Petar Sarić consists of two parts; in it are processed or present two wars, two major wars in the region of Montenegro and Herzegovina, the First and Second World War. However, it is more novel about divisions within the family and the man himself, (and infamous assault of godfather Luka on Sarah also and his murder are in that function, in the first part; and on the divisions among the people, in general, in the second part of the novel. The second part is, in fact, the image layering of life, not a symbolic one, full of hope, faith, reliance, rather than a concrete, real life, that life which is transformed into a fear of life. Separate, poetical, part of the novel, is his main character, Sara. It is no coincidence that her name novel entitled. Because she is one of most beautiful characters in the newer Serbian prose. Speech about the Sara precedes speech about her book. The book is Sara, Sara's book! Possession of book is her main feature of the exterior. Sara comes out from the Book and disappears in the book. Self contained and independent, therefore doomed to conflict with the environment. Loyal to husband and family, loyal to the truth and for justice, she ,,not hurt anything and anyone, no one is standing in the way, to anyone not wroth, nor has anyone looked wrong.' At the same time, the strange beauty, beauty that could not fit into some sort of scheme, one particular image or idea of beauty that again and again renewed, changed, remaining distant, and unmet. Strange goodness, marvelous beauty, she suffered unusual way; her life was transformed into continuous abstinence, repression, in anxiety and fear. In a word: in martirizam! Finally, in order to safe­guard children, sacrificed herself. Novel is a strong critique of society which is not able to recognize the beauty / goodness!.

Experimental and theoretical study of large scale debris bed reflood in the PEARL facility

Energy Technology Data Exchange (ETDEWEB)

Chikhi, Nourdine, E-mail: nourdine.chikhi@irsn.fr; Fichot, F.

2017-02-15

Highlights: • Five reflooding tests have been carried out with an experimental bed, 500 mm in height and 500 mm in diameter, made of 4 mm stainless steel balls. • For the first time, such a large bed was heated practically homogenously. • The quench front velocity was determined according to thermocouple measurements inside the bed. • An analytical model, assuming a quasi-steady progression of the quench front, allows to predict the conversion ratio in most cases. • It appears that the efficiency of cooling can be increased only up to a certain limit when increasing the inlet water flow rate. - Abstract: During a severe accident in a nuclear power plant, the degradation of fuel rods and melting of materials lead to the accumulation of core materials, which are commonly, called “debris beds”. To stop core degradation and avoid the reactor vessel rupture, the main accident management procedure consists in injecting water. In the case of debris bed, the reflooding models used for Loss of Coolant Accident are not applicable. The IRSN has launched an experimental program on debris bed reflooding to develop new models and to validate severe accident codes. The PEARL facility has been designed to perform, for the first time, the reflooding of large scale debris bed (Ø540 mm, h = 500 mm and 500 kg of steel debris) in a pressurized containment. The bed is heated by means of an induction system. A specific instrumentation has been developed to measure the debris bed temperature, pressure drop inside the bed and the steam flow rate during the reflooding. In this paper, the results of the first integral reflooding tests performed in the PEARL facility at atmospheric pressure up to 700 °C are presented. Focus is made on the quench front propagation and on the steam flow rate during reflooding. The effect of water injection flow rate, debris initial temperature and residual power are also discussed. Finally, an analytical model providing the steam flow rate and

Use of moving heat conductor mesh to perform reflood calculations with RELAP4/MOD6

International Nuclear Information System (INIS)

Fischer, S.R.; Ellis, L.V.; Chen, Y.S.

1979-01-01

RELAP4 is a computer code which can be used for the transient thermal hydraulic analysis of light water reactors and related systems. RELAP4/MOD6 includes many new analytical models which were developed primarily for the analysis of the reflood phase of a PWR loss-of-coolant accident (LOCA) transient. The key feature forming the basis for the MOD6 reflood calculation is a unique moving finite differenced heat conductor. The development and application of the moving heat conductor mesh for use in reflood analysis are described

Gene : CBRC-SARA-01-1996 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1996 Novel UN D UNKNOWN NU2C_ANTFO 0.004 26% gb|AAQ05281.1| NADH dehydrogenase subunit B [Metas...equoia glyptostroboides] 0.044 26% MILMLFLLGSSPRSLIHPRAILKTFPIVCCCLYAFCSLTSLILSLAVL

A 'quick-look' report on the THETIS 80% blocked cluster forced reflood experiments

International Nuclear Information System (INIS)

Cooper, C.A.; Pearson, K.G.

1984-01-01

A brief selection of results of forced reflooding experiments with the THETIS 80 percent blocked cluster is presented. A description of the THETIS blocked cluster test assemblies, and details of the test conditions, are given. The two forced reflooding experiments have been the subject of a blind calculation exercise with the BART code, and the results of these experiments are compared with the results from corresponding experiments with the 90 percent blocked cluster test assembly. Some general observations are made, arising from the comparison of these two series of experiments, and a qualitative explanation for the relatively complex variation of the heat transfer within the THETIS blockages is advanced. A full report on the 80 percent blocked cluster forced reflooding experiments will be available later. (U.K.)

Geometric effects of spacer grid in an annulus flow channel during reflooding period

International Nuclear Information System (INIS)

Cho, S.; Chun, S. Y.; Kim, B. D.; Park, J. K.; Yun, Y. J.; Baek, W. P.

2004-01-01

A number of studies on the reflooding phase were actively carried out from the early 70's due to its importance for the safety of the nuclear reactor. (Martini et al., 1973; Henry, 1974; Chung, 1978;) However, few studies have presented the spacer grid effect during the reflooding period. Since the grid is an obstruction in the flow passage, it causes an increased pressure drop due to form and skin friction losses. On the other hand, the spacer grid tends to increase the local wall heat transfer. The present work has been performed in a vertical annulus flow channel with various flow conditions. The objective of this paper is to evaluate the effects of a swirl-vane spacer grid on the rewetting phenomena during the reflooding phase

Staphylococcus aureus sarA regulates inflammation and colonization during central nervous system biofilm formation.

Directory of Open Access Journals (Sweden)

Jessica N Snowden

Full Text Available Infection is a frequent and serious complication following the treatment of hydrocephalus with CSF shunts, with limited therapeutic options because of biofilm formation along the catheter surface. Here we evaluated the possibility that the sarA regulatory locus engenders S. aureus more resistant to immune recognition in the central nervous system (CNS based on its reported ability to regulate biofilm formation. We utilized our established model of CNS catheter-associated infection, similar to CSF shunt infections seen in humans, to compare the kinetics of bacterial titers, cytokine production and inflammatory cell influx elicited by wild type S. aureus versus an isogenic sarA mutant. The sarA mutant was more rapidly cleared from infected catheters compared to its isogenic wild type strain. Consistent with this finding, several pro-inflammatory cytokines and chemokines, including IL-17, CXCL1, and IL-1β were significantly increased in the brain following infection with the sarA mutant versus wild type S. aureus, in agreement with the fact that the sarA mutant displayed impaired biofilm growth and favored a planktonic state. Neutrophil influx into the infected hemisphere was also increased in the animals infected with the sarA mutant compared to wild type bacteria. These changes were not attributable to extracellular protease activity, which is increased in the context of SarA mutation, since similar responses were observed between sarA and a sarA/protease mutant. Overall, these results demonstrate that sarA plays an important role in attenuating the inflammatory response during staphylococcal biofilm infection in the CNS via a mechanism that remains to be determined.

Experimental study of the reflooding of a constricted tube in the REFLEX rig

International Nuclear Information System (INIS)

Denham, M.K.; Elliott, D.F.; Britton-Jones, K.A.

1982-08-01

The Winfrith experimental programme in support of the PWR is focussed on fuel thermal and hydraulic performance under hypothetical accident conditions, and includes studies of reflooding heat transfer of single tubes and fuel rod clusters under simulated accident conditions, aimed at improving understanding of the processes involved and providing data for code development and validation. The work described is part of a study of the possible effects of clad ballooning on ECCS effectiveness. During a large loss of coolant accident the primary circuit will depressurise and the core will overheat. The Zircaloy fuel cladding may swell, partially blocking the coolant passages by the formation of local ''balloons''. An experiment was carried out in the REFLEX single tube reflooding rig, to study, in a simple geometry, the effect of the partial blockage of the tube on the fluid flow and heat transfer during reflooding. The blockage consisted of a tapering entrance with a flow area 60 percent less than the unconstricted tube, and a tapering exit. The flow could be viewed through windows. 66 refloods were carried out over a pressure range of 1 to 4 bar. Results of these tests are presented. (U.K.)

Evaluation report on CCTF Core-II reflood test C2-1 (Run 55)

International Nuclear Information System (INIS)

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

1991-10-01

A high pressure test (0.42 MPa) on the reflood phenomena was performed with the CCTF. The result of the test was compared with the experimental result of the base case test (0.2 MPa). (1) The overall flow characteristics in the high pressure test were qualitatively similar to that of the base case test. Any qualitatively different phenomena were not recognized during reflood phase. This indicates that it is reasonable to utilize the physical reflood model developed from the result of the base case test to the high pressure condition at least up to 0.42 MPa for prediction of reflood behavior of PWRs. (2) On the other hand, following quantitative influence of high pressure on reflood phenomena was observed. The core cooling was better, and the mass flow rate of the steam generated in the core was larger. However, the steam velocity was smaller due to higher density of the steam. Therefore, the steam discharge through loops was easier and hence the so-called steam binding effect was weaker. And, the water accumulation rate in the core was larger. Consequently the core flooding mass flow rate was larger. Since the core cooling was better, the maximum core temperature was lower and the last quenching was earlier. This result was the same as that previously observed in CCTF tests in the scope of the pressure upto 0.3 MPa. (3) The higher pressure leads to the better core cooling, and hence the safety margin increases with the increase in the pressure. (author)

Teologi Politik Berbalu SARA Antara Ambisi dan Konspirasi

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M. Sidi Ritau'din

2017-06-01

Full Text Available In a multi-cultural democracy based on Pancasila philosophy of independence, ethnic, religious, racial and intergroup issues it call (SARA are political indicators that can trigger conflict and division. If the player is ambitious and power-hungry, then he will not hesitate to do everything he can to gain power, even build a big conspiracy using SARA as a tool to divide the ummah, then he emerges as a unifier and presents programs prestigious sympathetic, there Imaging actions and slogans of the pro poor people, but essentially no more as political deceit, a false gift of hope, it familiar said (PHP that never realized, only reap the political advantage in the game of SARA, even not hesitate to shout thief when he Itself is a thieving thief based on greed and greed where the horizontal relations of fellow human beings deny the bond of faith as the foundation. Political conspiracy based on political interests and abuse of power, an action of political pathology that is not civilized that has become a trend of contemporary politics and globalization.

Calculations of flow oscillations during reflood using RELAP4/MOD6

International Nuclear Information System (INIS)

Chen, Y.S.; Fischer, S.R.; Sullivan, L.H.

1979-01-01

RELAP4/MOD6 is an analytical computer code which can be used for best-estimate analysis of LWR reactor system blowdown and reflood response to a postulated LOCA. In this study, flow oscillations in the PKL reflood test K5A were investigated using RELAP4/MOD6. Both calculated and measured oscillations exhibited transient characteristics of density-wave and pressure-drop oscillations. The calculated average core mixture level rising rate agrees closely with the test data. Several mechanisms which appear to be responsible for initiation and continuation of calculated or experimental reflood flow oscillations are (a) the coupling between the vapor generation in the core channel and the U-tube geometrical arrangement of a downcomer and a heated core; (b) the inherent low core inlet resistance and the high system outlet resistance; (c) the dependence of heat transfer rate on mass flow rate especially in the dispersed flow ially in the dispersed flow regime; (d) the amount of the liquid entrainment fraction of the heated core channel

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)

Rumen Microbiome Composition in Cattle during Grain-Induced Subacute Ruminal Acidosis (SARA)

DEFF Research Database (Denmark)

Danscher, Anne Mette; Derakshani, Hooman; Li, Shucong

2014-01-01

at the genus level. The rumen bacterial communities were altered in response to SARA (P=0.01). The proportion of several taxa was significantly higher in SARA samples, including S24-7, Erysipelotrichales. Lactobacillus, C lostridia, Moryella, Butyrivibrio, Olsenella, and C oprococcus. Microbiome profiling...

Prediction of reflood behavior for tests with differing axial power shapes using WCOBRA/TRAC

International Nuclear Information System (INIS)

Bajorek, S.M.; Hochreiter, L.E.

1991-01-01

The rector core power shape can vary over the fuel cycle due to load follow, control rod movement, burnup effects and Xenon transients. a best estimate thermal-hydraulic code must be able to accurately predict the reflooding behavior for different axial power shapes in order to find the power shapes effects on the loss-of-coolant peak cladding temperature. Several different reflood heat transfer experiments have been performed at the same or similar PWR reflood conditions with different axial power shapes. These experiments have different rod diameters, were full length, 3.65 m (12 feet) in height, and had simple egg crate grids. The WCOBRA/TRAC code has been used to model several different tests from these three experiments to examine the code's capability to predict the reflood transient for different power shapes, with a consistent model and noding scheme. This paper describes these different experiments, their power shapes, and the test conditions. The WCOBRA/TRAC code is described as well as the noding scheme, and the calculated results will be compared in detail with the test data rod temperatures. An overall assessment of the code's predictions of these experiments is presented

Code option guideline improvement using comparisons of RELAP4/MOD6 with forced and gravity-feed reflood data. [PWR

Energy Technology Data Exchange (ETDEWEB)

Chen, T H; Fletcher, C D

1978-09-01

Improved guidelines are developed for the selection of RELAP4/MOD6 reflood heat transfer options. The development, involving modifications to the original guidelines, assessed the effect of those modifications on RELAP4/MOD6 data comparisons using previously analyzed reflood experiments. The report also presents an evaluation of the application of the revised guidelines. Data comparisons between RELAP4/MOD6, using the original and revised guidelines, and experimental data are presented for Semiscale and FLECHT, forced-feed reflood tests and Semiscale and FLECHT-SET gravity-feed reflood tests. Because a general improvement was evident in data comparisons using the revised guidelines, their use is recommended in future calculations.

Study of the thermo-hydrodynamic phenomena in the nuclear core during reflood phase

International Nuclear Information System (INIS)

Murao, Yoshio

1983-03-01

This paper describes the development of the core thermo-hydrodynamic model on the reflood phenomena during a loss-of-coolant accident in a light water reactor. This model was developed based on the physical understanding in order to obtain the flexibility of application to safety analysis. For this purpose, the flow pattern was modeled and the fundamental equations were derived. The equations were used to know the suitable variables for assembling the thermo-hydrodynamic model of each flow regime in a reflood analysis code. Then the hydrodynamic models and the heat transfer models of all flow regimes and the quench model were derived. Some of them were newly developed. It was found that water accumulation above the quench front occurred in some cases, however the criteria was not clarified. One-dimensional forced-feed reflood tests were performed and the models were assessed and partly improved by using the data of the tests. The verified models were built in a one-dimensional reflood analysis code and totally assessed with the data of the test mentioned above. Except for the location just below a grid spacer and cases of high flooding rate, the calculational results indicated good comparison with the experimental results when the water accumulation was assumed above the quench front. Additionally the test data from the other test facility were used for the verification of the model. The results also showed good comparison with the experimental results. It was found that better comparisons were obtained when the water accumulation was not assumed above quench front. From these assessment of the model, it was found that the model derived here describes the over-all reflood phenomena, while it has to be partly improved and the water accumulation phenomena should be further investigated. (author)

ESTETIKA TARIAN SARA DOUDA DALAM MASYARAKAT ADAT LOLI (SEBUAH PENDEKATAN LINGUISTIK KEBUDAYAAN

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Sulistyastuti Sutomo

2015-01-01

Full Text Available Not only does art   have   self-fullfillment, but  it also has axiological  benefits both socially, culturally, religiously, and economiclally.  So does Sara Douda dance.  Sara Douda aesthetics  is first contained  in its whole dance movements. In addition, it  can also be found in the whole dance equipments.  Moreover,  this dance aesthetics  may  also be contained in the verbal symbols in the speech forms  prior to the dance performance. However,  both verbal and non-verbal aesthetical forms are incorporated by the pieces of socio-cultural and  religious values in the Loli community about  their honoring their ancestors,  having social harmony, and  highly respecting each other among the community members. This study uses a cultural linguistic approach to find out and to review the aesthetics of Sara Douda dance.   Seni memang memiliki kepenuhan dalam dirinya sendiri. Tetapi ia juga sekaligus punya faedah aksiologis, baik secara sosial, kultural, religius maupun secara ekonomis. Tarian Sara Douda pun demikian. Estetika Sara Douda pertama-tama ada dalam semua gerak tariannya. Juga dalam seluruh perlengkapan tarian tersebut. Bukan itu saja, estetika tarian ini juga ada dalam simbol-simbol verbal berupa tuturan menjelang tarian. Tetapi baik bentuk-bentuk estetisasi nonverbal maupun verbal, sama-sama disatukan oleh kepingan-kepingan nilai-nilai sosio-kultural dan religius masyarakat Loli tentang penghormatan kepada leluhur, tentang harmoni sosial, dan tentang penghargaan yang tinggi terhadap satu sama lain. Penelitian ini menggunakan pendekatan linguistik kebudayaan demi menemukan dan menelaah estetika dalam tarian Sara Douda

A Study on Uncertainty Quantification of Reflood Model using CIRCE Methodology

International Nuclear Information System (INIS)

Jeon, Seongsu; Hong, Soonjoon; Oh, Deogyeon; Bang, Youngseok

2013-01-01

The CIRCE method is intended to quantify the uncertainties of the correlations of a code. It may replace the expert judgment generally used. In this study, an uncertainty quantification of reflood model was performed using CIRCE methodology. In this paper, the application process of CIRCE methodology and main results are briefly described. This research is expected to be useful to improve the present audit calculation methodology, KINS-REM. In this study, an uncertainty quantification of reflood model was performed using CIRCE methodology. The application of CIRCE provided the satisfactory results. This research is expected to be useful to improve the present audit calculation methodology, KINS-REM

Fluiddynamic effects in the fuel element top nozzle area during refilling and reflooding

International Nuclear Information System (INIS)

Hawighorst, A.; Kroening, H.; Mewes, D.; Spatz, R.; Mayinger, F.

1985-01-01

During the refilling and reflooding phase following a hypothetical loss of coolant accident in lightwater cooled nuclear reactors, there will be countercurrent flow between discharging steam and the feed of emergency core cooling water. It was the objective of this research project to contribute to a better physical understanding of the fluiddynamic processes in the area of the fuel element top nozzle and so to improve emergency core cooling calculations. Therefore, experimental and theoretical investigations about the entrainment and countercurrent behaviour of gas/liquid flows have been implemented within this project. Fluiddynamic processes in the fuel element top nozzle area were simulated during the reflooding and refilling phase. Based on special internals as single and multiple-hole orifices, basic phenomena of fluidynamics were studied first with air-water. Subsequently, investigations of the system steam/water were conducted. The reactor geometry was approximated step by step, until a complete reactor fuel assembly top nozzle was constituted. The system pressure was 4.8 bars (abs), in accordance with the conditions in the reactor pressure vessel at the end of the blowdown phase. The water was initially fed in at saturation temperature, then, as a second step, fed in at subcooled condition relative to the steam temperature, in order to be able to study condensation effects as well. First, investigations on gas/liquid countercurrent flows in the fluid system air/water are presented. Then one studies countercurrent flow in the system steam/water, including the investigation of condensation effects. Finally, a detailed description of the research on droplet size determination is given

Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA

International Nuclear Information System (INIS)

Sugimoto, Jun

1989-01-01

This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)

Evaluation of the Trac-PF1 code for simulating the Neptun reflooding experiment

International Nuclear Information System (INIS)

Pontedeiro, A.C.; Galetti, M.R.S.

1991-01-01

The present work presents an assessment of the TRAC-BF1 code using the results of the NEPTUN experiment which simulates the reflooding in a loss-of-coolant accident (LOCA) in a PWR. The NEPTUN experiment is composed of an array of electrically-heated tubes where the reflooding condition can be tested. Two types of tests results are presented and compared with the values obtained with the TRAC-BF1 code. From this comparison it is concluded that TRAC is suitable for verifying accident analysis. (author)

System Analysis and Risk Assessment System (SARA), Version 4.0

International Nuclear Information System (INIS)

Russell, K.D.; Sattison, M.B.; Skinner, N.L.; Stewart, H.D.; Wood, S.T.

1992-02-01

This NUREG is the reference manual for the System Analysis and Risk Assessment (SARA) System Version 4.0, a microcomputer-based system used to analyze the safety issues of a family [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. The SARA data base contains PRA data for the dominant accident sequences of a family and descriptive information about the family including event trees, fault trees, and system model diagrams. The number of facility data bases that can be accessed is limited only by the amount of disk storage available. To simulate changes to family systems, SARA users change the failure rates of initiating and basic events and/or modify the structure of the cut sets that make up the event trees, fault trees, and systems. The user then evaluates the effects of these changes through the recalculation of the resultant accident sequence probabilities and importance measures. The results are displayed in tables and graphs

Konflik SARA pada Pilkada DKI Jakarta di Grup WhatsApp dengan Anggota Multikultural

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Tiara Kharisma

2017-12-01

Full Text Available Keanekaragaman SARA di Indonesia melahirkan masyarakat multikultural. Dalam kehidupannya, komunikasi antarbudaya tidak dapat dihindarkan. Salah satu medium dalam melakukan komunikasi antarbudaya adalah media sosial. Pada masyarakat multikultural isu SARA menjadi faktor utama penyebab terjadinya konflik. Di Pilkada DKI Jakarta 2017, isu SARA di grup WhatsApp marak menyebar termasuk anggota grup yang heterogen. Penelitian ini bertujuan untuk mengetahui pengelolaan konflik isu SARA pada Pilkada DKI Jakarta 2017 di grup WhatsApp dengan anggota multikultural. Penelitian ini menggunakan pendekatan kualitatif dengan teknik pengumpulan data melalui wawancara dan studi literatur. Dalam membahas penelitian ini, peneliti menggunakan kerangka teori manajemen konflik dari Martin, J. N. dan Nakayama. Hasil penelitian menunjukkan bahwa konflik terjadi karena ada anggota grup menyampaikan pesan bukan berangkat dari kesamaan anggota grup, yakni kepentingan dan tujuan awal dibentuknya grup. Pesan disebarkan dengan menganggap wujud pembelaan terhadap suatu agama. Ketika konflik terjadi, strategi pengelolaan konflik yang diterapkan adalah strategi mengompromikan (compromising dan menghindar (avoiding. Dalam grup terdapat anggota yang berperan sebagai cultural brokers.

Investigation of reflood models by coupling REFLA-1D and multi-loop system model

International Nuclear Information System (INIS)

Sugimoto, Jun; Murao, Yoshio

1983-09-01

A system analysis code REFLA-1DS was developed by coupling reflood analysis code REFLA-1D and a multi-loop primary system model. The reflood models in the code were investigated for the development of the integral system analysis code. The REFLA-1D, which was developed with the small scale reflood experiment at JAERI, consists of one-dimensional core model and a primary system model with a constant loop resistance. The multi-loop primary system model was developed with the Cylindrical Core Test Facility of JAERI's large scale reflood tests. The components modeled in the code are the upper plenum, the steam generator, the coolant pump, the ECC injection port, the downcomer and the broken cold leg nozzle. The coupling between the two models in REFLA-1DS is accomplished by applying the equivalent flow resistance calculated with the multiloop model to the REFLA-1D. The characteristics of the code is its simplicity of the system model and the solution method which enables the fast running and the easy reflood analysis for the further model development. A fairly good agreement was obtained with the results of the Cylindrical Core Test Facility for the calculated water levels in the downcomer, the core and the upper plenum. A qualitatively good agreement was obtained concerning the parametric effects of the system pressure, the ECC flow rate and the initial clad temperature. Needs for further code improvements of the models, however, were pointed out. These include the problem concerning the generation rate of the steam and water droplets in the core in an early period, the effect of the flow oscillation on the core cooling, the heat release from the downcomer wall, and the stable system calculation. (author)

ORNL instrumentation performance for Slab Core Test Facility (SCTF)-Core I Reflood Test Facility

International Nuclear Information System (INIS)

Hardy, J.E.; Hess, R.A.; Hylton, J.O.

1983-11-01

Instrumentation was developed for making measurements in experimental refill-reflood test facilities. These unique instrumentation systems were designed to survive the severe environmental conditions that exist during a simulated pressurized water reactor loss-of-coolant accident (LOCA). Measurement of in-vessel fluid phenomena such as two-phase flow velocity and void fraction and film thickness and film velocity are required for better understanding of reactor behavior during LOCAs. The Advanced Instrumentation for Reflood Studies (AIRS) Program fabricated and delivered instrumentation systems and data reduction software algorithms that allowed the above measurements to be made. Data produced by AIRS sensors during three experimental runs in the Japanese Slab Core Test Facility are presented. Although many of the sensors failed before any useful data could be obtained, the remaining probes gave encouraging and useful results. These results are the first of their kind produced during simulated refill-reflood stage of a LOCA near actual thermohydrodynamic conditions

Criterion for the onset of quench for low-flow reflood

International Nuclear Information System (INIS)

Hsu, Y.Y.; Young, M.W.

1982-07-01

This study provides a criterion for the onset of quench for low flow reflood. The criterion is a combination of two conditions: T/sub clad/ < T/sub limiting quench/ where T = Temperature, and α < 0.95 where α = void fraction. This criterion was obtained by examining temperature data from tests simulating PWR reflood, such as FLECHT, THTF, PBF, CCTF, and FEBA tests, with void fraction data from CCTF, FEBA, and FLECHT low flood tests. The data show that quenching initiated at α = 0.95 and that the majority of quench occurred at void fractions near 0.85. The results show that rods can be completely quenched by entrained droplets even if the collapsed liquid level does not advance. A thorough discussion of the analysis which supports this quench criterion is given in the text of this report

Study of top reflooding in case of severe accident and in particular oxidation of Uranium, Zirconium, Oxygen melts

International Nuclear Information System (INIS)

Brunet-Thibault, E.

2006-12-01

In 1979, the Three Mile Island (TMI) accident occurred in United States and accelerated research activities in the field of severe accidents. Severe accident management procedures imply massive water injections to flood the core. The work of this thesis bent principally over this reflooding. The first part of the study concerns the core oxidation enhancement during the reflooding phase which leads to a rough increase of the concentration of burnable hydrogen in the containment. This is why the study carried on the analysis of the contribution of the oxidation of U-Zr-O mixtures, towards the total production of hydrogen during reflooding. In the second part, the study concerns top flooding modelling i.e.: with injection of water in the hot legs. Here, we attempted to define bases and realize a model allowing to describe this type of reflooding. These models were validated on the simulation of the parameter with MAAP4 code. (author)

Effect of upper plenum water accumuration on reflooding phenomena under forced-feed flooding in SCTF Core-I tests

International Nuclear Information System (INIS)

Sudo, Yukio; Sobajima, Makoto; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1983-07-01

Large Scale Reflood Test Program has been performed under contract with the Atomic Energy Bureau of Science and Technology Agency of Japan since 1976. The Slab Core Test Program is a part of the Large Scale Reflood Test Program along with the Cylindrical Core Test Program. Major purpose of the Slab Core Test Program is to investigate two-dimensional, thermo-hydrodynamic behavior in the core and the effect of fluid communication between the core and the upper plenum on the reflood phenomena in a postulated loss-of-coolant accident of a PWR. A significant upper plenum water accumulation was observed in the Base Case Test Sl-01 which was carried out under forced-feed flooding condition. To investigate the effects of upper plenum water accumulation on reflooding phenomena, accumulated water is extracted out of the upper plenum in Test Sl-03 by full opening of valves for extraction lines located just above the upper core support plate. This report presents this effect of upper plenum water accumulation on reflooding phenomena through the comparison of Tests Sl-01 and Sl-03. In spite of full opening of valves for upper plenum water extraction in Test Sl-03, a little water accumulation was observed which is of the same magnitude as in Test Sl-01 for about 200 s after the beginning of reflood. From 200 s after the beginning of reflood, however, the upper plenum water accumulation is much less in Test Sl-03 than in Test Sl-01, showing the following effects of upper plenum water accumulation. In Test Sl-03, (1) the two-dimensionality of horizontal fluid distribution is much less both above and in the core, (2) water carryover through hot leg and water accumulation in the core are less, (3) quench time is rather delayed in the upper part of the core by less water fall back from the upper plenum, and (4) difference in the core thermal behavior and core heat transfer are not significant in the middle and lower part of the core. (author)

Safety Technology Research Program in the field of pressurized water reactors. 1. Technical report on advancement project RS 36/2. Emergency cooling program service life experiments: reflooding experiments involving the primary loop systems

International Nuclear Information System (INIS)

Schweickert, H.; Kremin, H.; Mandl, R.; Riedle, V.; Ruthrof, K.; Sarkar, J.; Schmidt, H.

The reflooding of the hot reactor core is to be examined for a pressurized water reactor (PWR), using a model of the entire primary loop system. The scale of the model is to be 1:340 in cross-section, with the heights represented full-scale. In addition to the goals of the project, a description of the test facility, including data collection and control equipment is presented. The instrumentation, the planned test program and the test procedure are briefly set forth

System Analysis and Risk Assessment (SARA) system

International Nuclear Information System (INIS)

Krantz, E.A.; Russell, K.D.; Stewart, H.D.; Van Siclen, V.S.

1986-01-01

Utilization of Probabilistic Risk Assessment (PRA) related information in the day-to-day operation of plant systems has, in the past, been impracticable due to the size of the computers needed to run PRA codes. This paper discusses a microcomputer-based database system which can greatly enhance the capability of operators or regulators to incorporate PRA methodologies into their routine decision making. This system is called the System Analysis and Risk Assessment (SARA) system. SARA was developed by EG and G Idaho, Inc. at the Idaho National Engineering Laboratory to facilitate the study of frequency and consequence analyses of accident sequences from a large number of light water reactors (LWRs) in this country. This information is being amassed by several studies sponsored by the United States Nuclear Regulatory Commission (USNRC). To meet the need of portability and accessibility, and to perform the variety of calculations necessary, it was felt that a microcomputer-based system would be most suitable

Experiment data report for semiscale MOD-1 tests S-03-A, S-03-B, S-03-C, and S-03-D (reflood heat transfer tests)

International Nuclear Information System (INIS)

1976-05-01

Recorded test data are presented for Tests S-03-A, S-03-B, S-03-C, and S-03-D of the Semiscale Mod-1 reflood heat transfer series (Test Series 3). The tests conducted in this series are separate effects core reflood tests performed to determine the reflood heat transfer characteristics of the 5.5 foot Mod-1 rod bundle. Tests S-03-A through S-03-D were forced-feed reflood tests in which the reflood rate was held constant during each test. The tests were conducted to investigate the effects on system response resulting from variations in operating conditions of pressure, temperature, core power, reflood coolant, subcooling, and peak heater rod thermocouple temperature at reflood initiation. Test S-03-A was conducted from an initial system temperature of about 230 0 F at a pressure of 20 psia. Tests S-03-B through S-03-D were conducted from an initial system temperature of about 290 0 F at a pressure of 60 psia. In all four tests, reflood coolant was injected directly into the core barrel by means of a specially designed core inlet manifold. The electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core during reflood. All four tests were conducted with a flat radial power profile. During reflood, core power was reduced from the initial level according to the American Nuclear Society (ANS) decay heat curve plus 20 percent for pressurized water reactor (PWR) core decay heat. The cold leg broken loop piping was open to the pressure suppression system (PSS). A separate steam supply system connected to the PSS was controlled to maintain constant pressure during the tests

Concurrent software system design supported by SARA at the age of one

Energy Technology Data Exchange (ETDEWEB)

Campos, I.M.; Estrin, G.

A multilevel modeling method suitable for the design of concurrent hardware or software systems is presented. The methodology is requirement driven and uses tools incorporated in a programming system called SARA (Systems ARchitect's Apprentice). Both top-down refinement and bottom-up abstraction are supported. The design of an asynchronous sender-receiver illustrates the key steps in going smoothly from programing in the large to programing in the small or actual code. The same methodology can be used to design hardware systems by applying different pragmatics from those proposed for software systems. SARA consists of a set of interactive tools implemented both at UCLA and also on the MIT-Multics system. Although SARA continues in long-term development, completed design tools are accessible for experimentation by authorized users at either location via the ARPANET. 9 figures, 2 tables.

Sensitive analysis and modifications to reflood-related constitutive models of RELAP5

International Nuclear Information System (INIS)

Li Dong; Liu Xiaojing; Yang Yanhua

2014-01-01

Previous system code calculation reveals that the cladding temperature is underestimated and quench front appears too early during reflood process. To find out the parameters shows important effect on the results, sensitive analysis is performed on parameters of constitutive physical models. Based on the phenomenological and theoretical analysis, four parameters are selected: wall to vapor film boiling heat transfer coefficient, wall to liquid film boiling heat transfer coefficient, dry wall interfacial friction coefficient and minimum droplet diameter. In order to improve the reflood simulation ability of RELAP5 code, the film boiling heat transfer model and dry wall interfacial friction model which are corresponding models of those influential parameters are studied. Modifications have been made and installed into RELAP5 code. Six tests of FEBA are simulated by RELAP5 to study the predictability of reflood-related physical models. A dispersed flow film boiling heat transfer (DFFB) model is applied when void fraction is above 0.9. And a factor is multiplied to the post-CHF drag coefficient to fit the experiment better. Finally, the six FEBA tests are calculated again so as to assess the modifications. Better results are obtained which prove the advantage of the modified models. (author)

Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

International Nuclear Information System (INIS)

Korteniemi, V.; Haapalehto, T.; Puustinen, M.

1995-01-01

Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied

Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

Energy Technology Data Exchange (ETDEWEB)

Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)

1995-09-01

Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

Cylindrical core reflood test facility (CCTF) and slab core reflood test facility (SCTF) for Japan Atomic Energy Research Institute (JAERI)

International Nuclear Information System (INIS)

1981-01-01

IHI has designed and constructed the CCTF at JAERI to be used in the safety analysis research on the loss of coolant accident in a PWR plant. This test facility is planned so that reflood phenomenon in the PWR plant (a phenomenon is that the bared and overheated core is reflooded by the emergency core cooling system when the coolant loss accident occurred) is simulated under various test conditions. The CCTF is the largest-scale test plant in the world, composed of approximately 2000 simulated fuel rods (electric heaters), 1 simulated pressure vessel, 4 primary cooling loops, 2 simulated steam generators, emergency core cooling system, and so on. The test conditions are controlled, and the test steps are sequentially progressed by the computing system, and test data are collected by the data acquisition system. Furthermore, IHI is now designing and constructing the SCTF in accordance with the JAERI research plan. The SCTF is similar to the CCTF in scale. Main feature of the SCTF is the form of the simulated core and the simulated pressure vessel, which is of slab construction to be representative of the radial section of the PWR reactor. Reliable and various data for safety analysis are expected by the CCTF and the SCTF. (author)

LOFT reflood as a function of accumulator initial gas volume

International Nuclear Information System (INIS)

Rhodes, H.F.

1978-01-01

The effect of the initial gas volume in the LOFT accumulators on the time to start of core reflood, after a LOCA, has been studied. The bases of the calculations are the data used and results presented in the Safety Analysis Report, Rev.1, August 1977, and the data in the RELAP and TOODEE2 program input and output listings. The results of this study show that an initial nitrogen volume of 12 cu ft, or more (at 600 psig initial pressure), would cause start of core reflood in time to prevent the cladding temperature from reaching 2200 0 F. The 12 cu ft initial volume will expand from 600 psig, initial pressure, to about 10 psig (containment pressure shortly after start of LOCA is approximately 8 psig) when all ECC liquid has been expelled from the accumulator. This pressure margin is considered too small; the ECC flowrate will be zero before the accumulator is empty

SARAS MEASUREMENT OF THE RADIO BACKGROUND AT LONG WAVELENGTHS

International Nuclear Information System (INIS)

Patra, Nipanjana; Subrahmanyan, Ravi; Sethi, Shiv; Shankar, N. Udaya; Raghunathan, A.

2015-01-01

SARAS is a correlation spectrometer connected to a frequency independent antenna that is purpose-designed for precision measurements of the radio background at long wavelengths. The design, calibration, and observing strategies admit solutions for the internal additive contributions to the radiometer response, and hence a separation of these contaminants from the antenna temperature. We present here a wideband measurement of the radio sky spectrum by SARAS that provides an accurate measurement of the absolute brightness and spectral index between 110 and 175 MHz. Accuracy in the measurement of absolute sky brightness is limited by systematic errors of magnitude 1.2%; errors in calibration and in the joint estimation of sky and system model parameters are relatively smaller. We use this wide-angle measurement of the sky brightness using the precision wide-band dipole antenna to provide an improved absolute calibration for the 150 MHz all-sky map of Landecker and Wielebinski: subtracting an offset of 21.4 K and scaling by a factor of 1.05 will reduce the overall offset error to 8 K (from 50 K) and scale error to 0.8% (from 5%). The SARAS measurement of the temperature spectral index is in the range −2.3 to −2.45 in the 110–175 MHz band and indicates that the region toward the Galactic bulge has a relatively flatter index

Improvement of MARS code reflood model

International Nuclear Information System (INIS)

Hwang, Moonkyu; Chung, Bub-Dong

2011-01-01

A specifically designed heat transfer model for the reflood process which normally occurs at low flow and low pressure was originally incorporated in the MARS code. The model is essentially identical to that of the RELAP5/MOD3.3 code. The model, however, is known to have under-estimated the peak cladding temperature (PCT) with earlier turn-over. In this study, the original MARS code reflood model is improved. Based on the extensive sensitivity studies for both hydraulic and wall heat transfer models, it is found that the dispersed flow film boiling (DFFB) wall heat transfer is the most influential process determining the PCT, whereas the interfacial drag model most affects the quenching time through the liquid carryover phenomenon. The model proposed by Bajorek and Young is incorporated for the DFFB wall heat transfer. Both space grid and droplet enhancement models are incorporated. Inverted annular film boiling (IAFB) is modeled by using the original PSI model of the code. The flow transition between the DFFB and IABF, is modeled using the TRACE code interpolation. A gas velocity threshold is also added to limit the top-down quenching effect. Assessment calculations are performed for the original and modified MARS codes for the Flecht-Seaset test and RBHT test. Improvements are observed in terms of the PCT and quenching time predictions in the Flecht-Seaset assessment. In case of the RBHT assessment, the improvement over the original MARS code is found marginal. A space grid effect, however, is clearly seen from the modified version of the MARS code. (author)

Droplet generation during core reflood

International Nuclear Information System (INIS)

Kocamustafaogullari, G.; De Jarlais, G.; Ishii, M.

1983-01-01

The process of entrainment and disintegration of liquid droplets by a flow of steam has considerable practical importance in calculating the effectivenes of the emergency core cooling system. Liquid entrainment is also important in determination of the critical heat flux point in general. Thus the analysis of the reflooding phase of a LOCA requires detailed knowledge of droplet size. Droplet size is mainly determined by the droplet generation mechanisms involved. To study these mechanisms, data generated in the PWR FLECHT SEASET series of experiments was analyzed. In addition, an experiment was performed in which the hydrodynamics of low quality post-CHF flow (inverted annular flow) were simulated in an adiabatic test section

Investigations of the reflood-phase after a loss-of-coolant-accident of an advanced pressurized water reactor (APWR)

International Nuclear Information System (INIS)

Schumann, S.; Oldekop, W.

1983-01-01

Differences between a high converting advanced pressurized-water reactor (APWR) and a conventional PWR, which are relevant to the reflood-phase after LOCA are presented. The used code and its verification by PWR-reflood experiments is explained. Comparative calculations for APWR and PWR with several conservative assumptions for example cold-leg-injection only, yield nearly the same maximum midplane-temperatures for the average-channel. For the APWR, however, the upper half of the rod shows higher temperatures. Quenchfront and core-water-level increase more slowly. The differences in the reflood-thermohydraulics are analysed in detail. A conservative hot-channel calculation shows maximum temperatures of about 920 0 C. Finally the influence of conservative assumptions is described and the necessity of experiments pointed out. (orig.)

RELAP5 model to simulate the thermal-hydraulic effects of grid spacers and cladding rupture during reflood

Energy Technology Data Exchange (ETDEWEB)

Nithianandan, C.K.; Klingenfus, J.A.; Reilly, S.S. [B& W Nuclear Technologies, Lynchburg, VA (United States)

1995-09-01

Droplet breakup at spacer grids and a cladding swelled and ruptured locations plays an important role in the cooling of nuclear fuel rods during the reflooding period of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). During the reflood phase, a spacer grid affects the thermal-hydraulic system behavior through increased turbulence, droplet breakup due to impact on grid straps, grid rewetting, and liquid holdup due to grid form losses. Recently, models to simulate spacer grid effects and blockage and rupture effects on system thermal hydraulics were added to the B&W Nuclear Technologies (BWNT) version of the RELAP5/MOD2 computer code. Several FLECHT-SEASET forced reflood tests, CCTF Tests C1-19 and C2-6, SCTF Test S3-15, and G2 Test 561 were simulated using RELAP5/MOD2-B&W to verify the applicability of the model at the cladding swelled and rupture locations. The results demonstrate the importance of modeling the thermal-hydraulic effects due to grids, and clad swelling and rupture to correctly predict the clad temperature response during the reflood phase of large break LOCA. The RELAP5 models and the test results are described in this paper.

Verification of thermal-hydraulic computer codes against standard problems for WWER reflooding

International Nuclear Information System (INIS)

Alexander D Efanov; Vladimir N Vinogradov; Victor V Sergeev; Oleg A Sudnitsyn

2005-01-01

Full text of publication follows: The computational assessment of reactor core components behavior under accident conditions is impossible without knowledge of the thermal-hydraulic processes occurring in this case. The adequacy of the results obtained using the computer codes to the real processes is verified by carrying out a number of standard problems. In 2000-2003, the fulfillment of three Russian standard problems on WWER core reflooding was arranged using the experiments on full-height electrically heated WWER 37-rod bundle model cooldown in regimes of bottom (SP-1), top (SP-2) and combined (SP-3) reflooding. The representatives from the eight MINATOM's organizations took part in this work, in the course of which the 'blind' and posttest calculations were performed using various versions of the RELAP5, ATHLET, CATHARE, COBRA-TF, TRAP, KORSAR computer codes. The paper presents a brief description of the test facility, test section, test scenarios and conditions as well as the basic results of computational analysis of the experiments. The analysis of the test data revealed a significantly non-one-dimensional nature of cooldown and rewetting of heater rods heated up to a high temperature in a model bundle. This was most pronounced at top and combined reflooding. The verification of the model reflooding computer codes showed that most of computer codes fairly predict the peak rod temperature and the time of bundle cooldown. The exception is provided by the results of calculations with the ATHLET and CATHARE codes. The nature and rate of rewetting front advance in the lower half of the bundle are fairly predicted practically by all computer codes. The disagreement between the calculations and experimental results for the upper half of the bundle is caused by the difficulties of computational simulation of multidimensional effects by 1-D computer codes. In this regard, a quasi-two-dimensional computer code COBRA-TF offers certain advantages. Overall, the closest

Mengukur Politisasi Agama dalam Ruang Publik: Komunikasi SARA dalam Perdebatan Rational Choice Theory

Directory of Open Access Journals (Sweden)

Mohammad Supriyadi

2015-12-01

Full Text Available Tulisan ini memberikan gambaran runtuhnya pengaruh isu primordialisme di ruang publik dan digantikan dengan kearifan konvensional. Penelitian ini mengambil aspek pengaruh isu SARA pada aspek rasionalitas pemilih. Penulis menemukan beberapa aspek yang mendukung kesimpulan penelitian, antara lain; bahwa isu SARA tidak terlalu direspek pemilih rasional. Pemilih rasional lebih melihat masalah yang ada dan mengevaluasi kinerja pemerintahan sebelumnya. Di lain pihak, emosi antusias terhadap isu etnisitas akan memantabkan pilihan politik terhadap pemilih etnis minoritas, sebagai bentuk penguatan komunitas. Dengan menggunakan pendekatan teori pilihan rasional (rational choice theory, penulis melihat bahwa komunikasi politik yang dibangun melalui isu SARA di ruang publik dalam kehidupan masyarakat modern, tidak lagi mampu memengaruhi pemilih rasional. Pemilih rasional (rational choice, menentukan pilihan berdasarkan pada keuntungan yang diperolehnya (maximizing benefit. Dalam faktor ini sikap pemilih lebih dipengaruhi karakteristik dan track record kandidat.

NCBI nr-aa BLAST: CBRC-SARA-01-1488 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1488 ref|ZP_01187451.1| conserved hypothetical protein [Bacillus weihenstep...hanensis KBAB4] gb|EAR73176.1| conserved hypothetical protein [Bacillus weihenstephanensis KBAB4] ZP_01187451.1 0.78 25% ...

NCBI nr-aa BLAST: CBRC-SARA-01-1089 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1089 ref|ZP_00518636.1| Major facilitator superfamily [Crocosphaera watson...ii WH 8501] gb|EAM48279.1| Major facilitator superfamily [Crocosphaera watsonii WH 8501] ZP_00518636.1 1.3 32% ...

NCBI nr-aa BLAST: CBRC-SARA-01-0489 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0489 ref|ZP_01035526.1| exopolysaccharide biosynthesis domain protein [Rose...ovarius sp. 217] gb|EAQ25691.1| exopolysaccharide biosynthesis domain protein [Roseovarius sp. 217] ZP_01035526.1 1.0 26% ...

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)

NCBI nr-aa BLAST: CBRC-SARA-01-1419 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1419 ref|YP_101126.1| hypothetical protein BF3850 [Bacteroides fragili...s YCH46] dbj|BAD50592.1| hypothetical protein [Bacteroides fragilis YCH46] YP_101126.1 5e-09 92% ...

NCBI nr-aa BLAST: CBRC-SARA-01-0522 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0522 ref|YP_001499570.1| Peptidyl-tRNA hydrolase [Rickettsia massiliae... MTU5] gb|ABV85023.1| Peptidyl-tRNA hydrolase [Rickettsia massiliae MTU5] YP_001499570.1 4.7 36% ...

NCBI nr-aa BLAST: CBRC-SARA-01-0900 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0900 ref|YP_898010.1| phosphatidylserine synthase [Francisella tularen...sis subsp. novicida U112] gb|ABK89256.1| phosphatidylserine synthase [Francisella tularensis subsp. novicida U112] YP_898010.1 3.0 27% ...

NCBI nr-aa BLAST: CBRC-SARA-01-0144 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0144 ref|YP_438659.1| alcohol dehydrogenase, zinc-containing [Burkholderia thailand...ensis E264] gb|ABC35120.1| alcohol dehydrogenase, zinc-containing [Burkholderia thailandensis E264] YP_438659.1 0.69 35% ...

NCBI nr-aa BLAST: CBRC-SARA-01-1967 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1967 ref|ZP_00370594.1| Predicted DNA repair exonuclease [Campylobacter upsaliens...is RM3195] gb|EAL53370.1| Predicted DNA repair exonuclease [Campylobacter upsaliensis RM3195] ZP_00370594.1 0.047 31% ...

NCBI nr-aa BLAST: CBRC-SARA-01-1753 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1753 ref|YP_001476411.1| amino acid permease-associated region [Serratia proteam...aculans 568] gb|ABV39283.1| amino acid permease-associated region [Serratia proteamaculans 568] YP_001476411.1 0.83 35% ...

Evaluation report on CCTF Core-II reflood tests C2-AC1 (run 51) and C2-4 (run 62)

International Nuclear Information System (INIS)

Sugimoto, Jun; Iguchi, Tadashi; Murao, Yoshio

1984-02-01

A reflood test program has been conducted at Japan Atomic Energy Research Institute (JAERI) using large scale test facilities named Cylindrical Core Test Facility (CCTF) and Slab Core Test Facility (SCTF). The present report describes the effect of the initial clad temperature i.e., the initial stored energy on reflood phenomena observed in CCTF Core-II tests C2-ACl and C2-4. The peak clad temperatures of tests C2-ACl and C2-4 were 863 K and 1069 K, respectively at reflood initiation. With higher initial clad temperature, obtained were lower water accumulation in the core and upper plenum, and higher loop mass flow rate in an early reflood transient due to larger heat release of the stored energy in the core. Core inlet flow conditions were only affected shortly after the reflood initiation, causing the suppressed flooding rate and the larger U-tube flow oscillation between the core and the downcomer. In the core, with higher initial clad temperature, slower quench front propagation and higher turnaround temperature were observed. Responses to a higher initial clad temperature were similar to those observed in CCTF Core-I and FLECHT tests. Thus, the lower temperature rise with higher initial clad temperature was experimentally confirmed. The importance of higher flooding rate at initial period was analytically shown for further decreasing the temperature rise. (author)

NCBI nr-aa BLAST: CBRC-SARA-01-1309 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1309 ref|YP_001125845.1| hypothetical protein GTNG_1736 [Geobacillus thermoden...itrificans NG80-2] gb|ABO67100.1| Conserved hypothetical protein [Geobacillus thermodenitrificans NG80-2] YP_001125845.1 0.47 23% ...

NCBI nr-aa BLAST: CBRC-SARA-01-0488 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0488 ref|YP_064024.1| sulfate permease (SulP) [Desulfotalea psychrophi...la LSv54] emb|CAG35017.1| probable sulfate permease (SulP) [Desulfotalea psychrophila LSv54] YP_064024.1 6.1 37% ...

NCBI nr-aa BLAST: CBRC-SARA-01-0771 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0771 ref|NP_001029253.1| progestin and adipoQ receptor family member V...II [Rattus norvegicus] gb|AAY67652.1| progestin membrane receptor alpha [Rattus norvegicus] NP_001029253.1 1e-167 82% ...

Safety verification for the ECCS driven by the electrically 4 trains during LBLOCA reflood phase using ATLAS

International Nuclear Information System (INIS)

Park, Yusun; Park, Hyun-sik; Kang, Kyoung-ho; Choi, Nam-hyun; Min, Kyoung-ho; Choi, Ki-yong

2014-01-01

Highlights: • Safety improvement by adopting 4 train emergency core cooling system was validated experimentally. • General thermal hydraulic behaviors of the system during LBLOCA reflood phase were successfully demonstrated. • Key parameters such as the liquid levels, the PCTs, the quenching time, and the ECC bypass ratios were investigated. • Asymmetric effects of the different combination of safety injection were negligible during the reflood period. - Abstract: The APR1400 is equipped with four safety injection pumps driven by two emergency diesel generators. However, the design has been changed so that the four safety injection pumps are driven by 4 emergency diesel generators during the design certification process from the U.S. NRC. Thus, 4 safety injection pumps (SIPs) are completely independent electrically and mechanically and three safety injection pumps are available in a single failure condition. This design change could have a certain effects on the thermal-hydraulic phenomenon occurring in the downcomer region during the late reflood phase of a large break loss of coolant accident (LBLOCA). Thus, in this study, a verification experiment for the reflood phase of a LBLOCA was performed to evaluate the core cooling performance of the 4 train emergency core cooling system (ECCS) with an assumption of a single failure. And the different combinations of three SIPs positions were tested to investigate the asymmetric effects on the reactor core cooling performance. The overall experimental results revealed the typical thermal–hydraulic trends expected to occur during the reflood phase of a large-break LOCA scenario for the APR1400. Experiment with the injection of three SIPs showed a faster core quenching time and lower bypass ratio than that of the case in which two SIPs were injected. The RPV wall temperature distributions showed the similar trend in spite of the different SIP combinations

Identification of Differentially Expressed Proteins in Liver in Response to Subacute Ruminal Acidosis (SARA Induced by High-concentrate Diet

Directory of Open Access Journals (Sweden)

X. Y. Jiang

2014-08-01

Full Text Available The aim of this study was to evaluate protein expression patterns of liver in response to subacute ruminal acidosis (SARA induced by high-concentrate diet. Sixteen healthy mid-lactating goats were randomly divided into 2 groups and fed either a high-forage (HF diet or a high-concentrate (HC diet. The HC diet was expected to induce SARA. After ensuring the occurrence of SARA, liver samples were collected. Proteome analysis with differential in gel electrophoresis technology revealed that, 15 proteins were significantly modulated in liver in a comparison between HF and HC-fed goats. These proteins were found mainly associated with metabolism and energy transfer after identified by matrix-assisted laser desorption ionization/time of flight. The results indicated that glucose, lipid and protein catabolism could be enhanced when SARA occurred. It prompted that glucose, lipid and amine acid in the liver mainly participated in oxidation and energy supply when SARA occurred, which possibly consumed more precursors involved in milk protein and milk fat synthesis. These results suggest new candidate proteins that may contribute to a better understanding of the mechanisms that mediate liver adaptation to SARA.

NCBI nr-aa BLAST: CBRC-SARA-01-1323 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1323 ref|NP_001008277.1| rhodopsin [Canis lupus familiaris] ref|XP_855...608.1| PREDICTED: rhodopsin [Canis familiaris] emb|CAA70209.1| unnamed protein product [Canis familiaris] NP_001008277.1 2e-62 94% ...

NCBI nr-aa BLAST: CBRC-SARA-01-0322 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0322 ref|ZP_01459778.1| conserved hypothetical protein [Stigmatella au...rantiaca DW4/3-1] gb|EAU69359.1| conserved hypothetical protein [Stigmatella aurantiaca DW4/3-1] ZP_01459778.1 5.2 34% ...

NCBI nr-aa BLAST: CBRC-SARA-01-1608 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1608 ref|XP_513201.1| PREDICTED: cannabinoid receptor 2 (macrophage) i...soform 2 [Pan troglodytes] ref|XP_001166334.1| PREDICTED: cannabinoid receptor 2 (macrophage) isoform 1 [Pan troglodytes] XP_513201.1 1e-155 77% ...

NCBI nr-aa BLAST: CBRC-SARA-01-0256 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0256 ref|ZP_00739965.1| Integral membrane protein [Bacillus thuringiensis serovar israel...ensis ATCC 35646] gb|EAO55770.1| Integral membrane protein [Bacillus thuringiensis serovar israelensis ATCC 35646] ZP_00739965.1 1.3 24% ...

Implementation of JAERI's reflood model into TRAC-PF1/MOD1 code

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

1993-02-01

Selected physical models of REFLA code, that is a reflood analysis code developed at JAERI, were implemented into the TRAC-PF1/MOD1 code in order to improve the predictive capability of the TRAC-PF1/MOD1 code for the core thermal hydraulic behaviors during the reflood phase in a PWR LOCA. Through comparisons of physical models between both codes, (1) Murao-Iguchi void fraction correlation, (2) the drag coefficient correlation acting to drops, (3) the correlation for wall heat transfer coefficient in the film boiling regime, (4) the quench velocity correlation and (5) heat transfer correlations for the dispersed flow regime were selected from the REFLA code to be implemented into the TRAC-PF1/MOD1 code. A method for the transformation of the void fraction correlation to the equivalent interfacial friction model was developed and the effect of the transformation method on the stability of the solution was discussed. Through assessment calculation using data from CCTF (Cylindrical Core Test Facility) flat power test, it was confirmed that the predictive capability of the TRAC code for the core thermal hydraulic behaviors during the reflood can be improved by the implementation of selected physical models of the REFLA code. Several user guidelines for the modified TRAC code were proposed based on the sensitivity studies on fluid cell number in the hydraulic calculation and on node number and effect of axial heat conduction in the heat conduction calculation of fuel rod. (author)

Simulation of reflooding on two parallel heated channel by TRACE

Energy Technology Data Exchange (ETDEWEB)

Zakir, Md. Ghulam [Department of Nuclear Engineering, Chalmers University of Technology, Gothenburg (Sweden)

2016-07-12

In case of Loss-Of-Coolant accident (LOCA) in a Boiling Water Reactor (BWR), heat generated in the nuclear fuel is not adequately removed because of the decrease of the coolant mass flow rate in the reactor core. This fact leads to an increase of the fuel temperature that can cause damage to the core and leakage of the radioactive fission products. In order to reflood the core and to discontinue the increase of temperature, an Emergency Core Cooling System (ECCS) delivers water under this kind of conditions. This study is an investigation of how the power distribution between two channels can affect the process of reflooding when the emergency water is injected from the top of the channels. The peak cladding temperature (PCT) on LOCA transient for different axial level is determined as well. A thermal-hydraulic system code TRACE has been used. A TRACE model of the two heated channels has been developed, and three hypothetical cases with different power distributions have been studied. Later, a comparison between a simulated and experimental data has been shown as well.

Report on series 4 reflood experiment

International Nuclear Information System (INIS)

Murao, Yoshio; Iguchi, Tadashi; Sudoh, Takashi; Sudo, Yukio; Sugimoto, Jun

1977-03-01

Series 4 reflood experiment was carried out from June to July 1976. The purpose was to examine system-pressure effect and system effect with heater rods having about the same heat capacity as the real ones and flow resistance of the primary loop. The results are: 1) The core flood velocity increases and the quench time decreases with system pressure. 2) The relation between core differential pressure (accumulation head of the core) and core power density, which indicates coolability of the core, is obtainable in map with the system pressure as a parameter. 3) The steady-state thermal condition occurs in the test section with constant power density. (auth.)

Reflooding phase after loss of coolant of an advanced pressurized water reactor with high conversion ratio

International Nuclear Information System (INIS)

Schumann, S.

1984-01-01

The emergency core cooling behaviour of an advanced pressurized water reactor (APWR) during the reflooding phase of the LOCA with double-ended break is analysed and compared to a common pressurized water reactor (PWR). The code FLUT-BS, its models and correlations are explained in detail and have been verified by numerous PWR-reflood experiments with large parameter range. The influence of core-design on ECC-behaviour as well as the influences of initial and boundary values are examined. The results show the essential differences of ECC-behaviour between PWR and APWR. (orig.) [de

NCBI nr-aa BLAST: CBRC-SARA-01-1099 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1099 ref|YP_001337495.1| hypothetical protein KPN_03841 [Klebsiella pneumoniae subsp. pneumonia...e MGH 78578] gb|ABR79228.1| hypothetical protein KPN_03841 [Klebsiella pneumoniae subsp. pneumoniae MGH 78578] YP_001337495.1 0.93 29% ...

NCBI nr-aa BLAST: CBRC-SARA-01-0771 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0771 ref|NP_001033642.1| progestin and adipoQ receptor family member V...II [Bos taurus] gb|AAI11285.1| Progestin and adipoQ receptor family member VII [Bos taurus] NP_001033642.1 1e-180 87% ...

Experimental study of effect of initial clad temperature on reflood phenomena during PWR-LOCA

International Nuclear Information System (INIS)

Sugimoto, Jun; Murao, Yoshio

1983-01-01

Integral system tests with the Cylindrical Core Test Facility (CCTF) were performed to investigate the effect of the initial clad temperature on the reflood phenomena in a PWR-LOCA. The initial peak clad temperatures in these three tests were 871, 968 and 1,047K, respectively. The feedback of the system on the core inlet mass flow rate was estimated to be little influenced by the variation of the initial clad temperature except for the first 20s in the transient. The observed temperature rise from the reflood initiation was lower with the higher initial clad temperature. This qualitatively agreed with the results of the small scale forced feed reflood experiments. However, the magnitude of the temperature rise in CCTF was significantly low due to the high initial core inlet mass flow rate. Also observed were the multi-dimensional thermal behaviors for the three cases in the CCTF wide core. The analysis codes REFLA and TRAC reasonably predicted the effect of the initial clad temperature on the core thermo-hydraulics under the simulated core inlet flow conditions. However, the calculated temperature rise of the maximum powered rod based on the one-dimensional core analysis was higher than that of the average powered rod, which contradicts the tendency observed in CCTF tests. (author)

System pressure effects on reflooding phenomena observed in the SCTF Core-I forced flooding tests

International Nuclear Information System (INIS)

Adachi, Hiromichi; Sudo, Yukio; Sobajima, Makoto; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka

1983-06-01

The Slab Core Test Facility was constructed to investigate two-dimensional thermo-hydrodynamics in the core and the interaction in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a posturated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report described the analytical results on the effects of system pressure on reflooding phenomena observed in Tests Sl-SH2, Sl-01 and Sl-02 which are belonging to the SCTF Core-I forced-feed reflooding test series. Nominal system pressures in these tests are 0.4, 0.2 and 0.15 MPa, respectively. By comparison among the data of these three tests, the effects of system pressure on thermo-hydrodynamic behavior in the pressure vessel including the core and the primary coolant loops of the SCTF can be clarified under the forced flooding condition. Major items investigated in the present report are (1) overall temperature behaviors in the core, (2) change of heat transfer coefficient and heat flux at the rod surface before the quench, (3) two-dimensional thermo-hydrodynamic behaviors in the core and upper plenum and (4) hot leg carryover. (author)

NCBI nr-aa BLAST: CBRC-SARA-01-1273 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1273 ref|ZP_00367235.1| competence locus E (comE3), putative [Campylob...acter coli RM2228] gb|EAL57139.1| competence locus E (comE3), putative [Campylobacter coli RM2228] ZP_00367235.1 0.001 28% ...

NCBI nr-aa BLAST: CBRC-SARA-01-1804 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1804 ref|ZP_00743402.1| Sensory box/GGDEF family protein [Bacillus thuringiensis serovar israel...ensis ATCC 35646] gb|EAO52327.1| Sensory box/GGDEF family protein [Bacillus thuringiensis serovar israelensis ATCC 35646] ZP_00743402.1 0.20 36% ...

Contribution to the modelling of flows and heat transfers during the reflooding phase of a PWR core

International Nuclear Information System (INIS)

Colas, D.

1984-01-01

This thesis contributes to modelise thermohydraulic phenomena occuring in a pressurized water nuclear reactor core during the reflood phase of a LOCA. The reference accident and phenomena occuring during reflooding are described as well as flow regime and heat transfer proposed models. With these models, we developed a code to compute fluid conditions and fuel rods temperatures in a reactor core chanel. In order to test this code, results of computation are compared with experiments (FLECHT Skewed Tests) and a conclusion is drawn [fr

Modeling of reflood of severely damaged reactor core

International Nuclear Information System (INIS)

Bachrata, A.

2012-01-01

The TMI-2 accident and recently Fukushima accident demonstrated that the nuclear safety philosophy has to cover accident sequences involving massive core melt in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although severe accidents are low likelihood and might be caused only by multiple failures, accident management is implemented for controlling their course and mitigating their consequences. In case of severe accident, the fuel rods may be severely damaged and oxidized. Finally, they collapse and form a debris bed on core support plate. Removal of decay heat from a damaged core is a challenging issue because of the difficulty for water to penetrate inside a porous medium. The reflooding (injection of water into core) may be applied only if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ from original rod bundle geometry and will resemble to the severe damaged core observed in TMI-2. The higher temperatures and smaller hydraulic diameters in a porous medium make the coolability more difficult than for intact fuel rods under typical loss of coolant accident conditions. The modeling of this kind of hydraulic and heat transfer is a one of key objectives of this. At IRSN, part of the studies is realized using an European thermo-hydraulic computer code for severe accident analysis ICARE-CATHARE. The objective of this thesis is to develop a 3D reflood model (implemented into ICARE-CATHARE) that is able to treat different configurations of degraded core in a case of severe accident. The proposed model is characterized by treating of non-equilibrium thermal between the solid, liquid and gas phase. It includes also two momentum balance equations. The model is based on a previously developed model but is improved in order to take into account intense boiling regimes (in particular

NCBI nr-aa BLAST: CBRC-SARA-01-1746 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1746 ref|NP_001100989.1| non imprinted in Prader-Willi/Angelman syndro...me 1 homolog [Rattus norvegicus] gb|EDL86455.1| non imprinted in Prader-Willi/Angelman syndrome 1 homolog (human) (predicted) [Rattus norvegicus] NP_001100989.1 1e-113 80% ...

NCBI nr-aa BLAST: CBRC-SARA-01-1578 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1578 ref|NP_001077100.1| cOR52N9 olfactory receptor family 52 subfamily N-like [Canis lupus... familiaris] gb|ABO36681.1| 52N9 olfactory receptor protein [Canis lupus familiaris] NP_001077100.1 1e-146 78% ...

Analysis of the core reflooding of a PWR reactor under a loss-of-coolant postulated accident

International Nuclear Information System (INIS)

Austregesilo Filho, H.

1978-12-01

The main purpose of this work is to analyse the termohydraulic behaviour of emergency cooling water, during reflooding of a PWR core submitted to a postulated loss-of-coolant accident, with the scope of giving the boundary conditions needed to verify fuel element and containment integrity. The analytical model presented was applied to the simulation of Angra I core reflooding phase, after a double-ended break between pressure vessel and discharge of one of the main coolant pumps. For this accident, with a discharge coefficient of C sub(D) = 0.4, the highest peak cladding temperature is expected. (author) [pt

Reactivity of Athabasca residue and of its SARA fractions during residue hydroconversion

Energy Technology Data Exchange (ETDEWEB)

Verstraete, J.; Danial-Fortain, P.; Gauthier, T.; Merdrignac, I. [IFP-Lyon, Vermaison (France); Budzinski, H. [Bordeaux Univ. (France). ISM-LPTC, UMR CNRS

2009-07-01

Residue conversion processes are becoming increasingly important because of the declining market for residual fuel oil and a greater demand for middle distillates. Ebullated-bed hydroconversion is a commercially proven technology for converting heavy feedstocks with high amounts of impurities. The process enables the conversion of atmospheric or vacuum residues at temperatures up to 440 degrees C, and at liquid hourly space velocity (LHSV) conditions in the range of 0.15 to 0.5 per hour. A 540 degrees C conversion of up to 80 weight per cent can be achieved under these conditions. This paper reported on a research study conducted at IFP Lyon in which the residue hydroconversion in a large-scale ebullated bed bench unit was investigated to determine the impact of operating conditions and feed properties on yield and product qualities. Hydrogen was added to the feed in the bench units to keep a high hydrogen partial pressure and favour the catalytic hydroconversion reactions. In a typical test, the reactor was fed with 50 g of feedstock and 0.45 g of crushed equilibrium industrial NiMo catalyst, pressurized hydrogen and quickly heated at the reaction temperature. This paper also discussed the conversion of Athabasca bitumen residue in the large-scale pilot plant and also in the small scale batch reactor. The effect of operating temperature and space velocity was examined. The reactivity of the saturates, aromatics, resins and asphaltenes (SARA) fractions of the bitumen was studied separately in order to better understand the conversion mechanisms and reactivities. The Athabasca bitumen feed and SARA fractions were also analyzed in terms of standard petroleum analysis, SARA fractionation, elemental analysis, size exclusion chromatography (SEC) and 13C NMR. Hydroconversion experiments were conducted in the batch unit at different reaction temperatures and reaction times. A comparison of small-scale batch results with those obtained with the continuous large-scale bench

Calculations of film boiling heat transfer above the quench front during reflooding

International Nuclear Information System (INIS)

Chan, K.C.; Yadigaroglu, G.

1980-01-01

An analytical method for calculating inverted-annular film boiling heat transfer above the quench front during the reflooding phase of a LOCA is presented. A two-fluid model comprising a laminar vapor film and a turbulent liquid-vapor mixture core is used. 12 refs

Impacts of Mesopotamian wetland re-flooding on the lipid biomarker distributions in sediments

Science.gov (United States)

Rushdi, Ahmed I.; DouAbul, Ali A. Z.; Al-Maarofi, Sama S.; Simoneit, Bernd R. T.

2018-03-01

Shallow sediment core samples from two locales in the Mesopotamian marshlands of Iraq were analyzed to characterize the extractable organic (lipid) compounds, and their sources and distributions after hydrological restoration by re-flooding of the marshes. Dried samples were extracted with a dichloromethane/methanol mixture before analysis by gas chromatography-mass spectrometry (GC-MS). The major compounds were n-alkanes, fatty acids and alcohols, steroids, terpenoids, hopanes, steranes, unresolved complex mixture (UCM), and plasticizers. The lipid compounds in Kurmashia (Al-Hammar marshes) were generally higher in concentration than in Abu Zirig (Central marshes), and decreased with core depths for both sites. This concentration decrease with core depth is attributed to transformation, biodegradation and variable input processes. The distribution patterns of the lipids in the sediment cores indicated that the Abu Zirig area was drier than Kurmashia before the re-flooding process. Furthermore, the concentration of the compounds in the surface sediment the Abu Zirig core was as high and similar to that in Kurmashia, reflecting the re-flooding impacts on the marsh and the revival of the wetland. The major sources of these lipids were from natural terrestrial vegetation (35-66% for Abu Zirig; 40-49% for Kurmashia), microbial (plankton) residues and bacteria (27-52% for Abu Zirig; 39-43% for Kurmashia), with a minor contribution from anthropogenic sources including plastic wastes and petroleum (6-13% for Abu Zirig; 9-18% for Kurmashia).

Calculations for the design and modification of the 2 cyclotrons of S.A.R.A

International Nuclear Information System (INIS)

Albrand, P.S.; Belmont, J.L.; Ripouteau, F.

1983-09-01

S.A.R.A. is a heavy ion accelerator constituted by 2 cyclotrons. The second cyclotron (post-accelerator) was entirely calculated at the I.S.N. The pole tips of the first cyclotron which is much older, have recently been modified. An almost identical procedure was used for the calculation of each element of the post-accelerator of S.A.R.A. and also for the modifications to the first cyclotron

Inhibition of transforming growth factor-beta1-induced signaling and epithelial-to-mesenchymal transition by the Smad-binding peptide aptamer Trx-SARA.

Science.gov (United States)

Zhao, Bryan M; Hoffmann, F Michael

2006-09-01

Overexpression of the inhibitory Smad, Smad7, is used frequently to implicate the Smad pathway in cellular responses to transforming growth factor beta (TGF-beta) signaling; however, Smad7 regulates several other proteins, including Cdc42, p38MAPK, and beta-catenin. We report an alternative approach for more specifically disrupting Smad-dependent signaling using a peptide aptamer, Trx-SARA, which comprises a rigid scaffold, the Escherichia coli thioredoxin A protein (Trx), displaying a constrained 56-amino acid Smad-binding motif from the Smad anchor for receptor activation (SARA) protein. Trx-SARA bound specifically to Smad2 and Smad3 and inhibited both TGF-beta-induced reporter gene expression and epithelial-to-mesenchymal transition in NMuMG murine mammary epithelial cells. In contrast to Smad7, Trx-SARA had no effect on the Smad2 or 3 phosphorylation levels induced by TGF-beta1. Trx-SARA was primarily localized to the nucleus and perturbed the normal cytoplasmic localization of Smad2 and 3 to a nuclear localization in the absence of TGF-beta1, consistent with reduced Smad nuclear export. The key mode of action of Trx-SARA was to reduce the level of Smad2 and Smad3 in complex with Smad4 after TGF-beta1 stimulation, a mechanism of action consistent with the preferential binding of SARA to monomeric Smad protein and Trx-SARA-mediated disruption of active Smad complexes.

NEPTUN/5052, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test

International Nuclear Information System (INIS)

Richner, M.; Analytis, G.Th.; Aksan, S.N.

1993-01-01

1 - Description of test facility: NEPTUN is designed to perform PWR LOCA simulation experiments, which provide the full length emergency cooling heat transfer tests for LOFT. Therefore the NEPTUN heater bundle with 33 electrical heater elements and 4 guide tubes simulates a section of the LOFT nuclear core. The main test loop also contains measuring systems for the carry-over rate and for the steam expelled, and a back-pressure control system. A water loop brings the water to the initial reflooding conditions. In addition, auxiliary systems maintain normal operating conditions. 2 - Description of test: Test 5052 is one of a series of 40 reflood tests performed in NEPTUN. Before the start of the test, the flooding water in its circuit is brought to the following conditions: pressure = 4.1 bar; velocity = 2.5 cm/sec; subcooling temperature = 78 C; single rod power = 2.45 kW; maximal initial cladding temperature = 867 C. 3 - Status: CSNI1013/01, 21-Jul-1993 Arrived at NEADB

NCBI nr-aa BLAST: CBRC-SARA-01-1488 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1488 ref|YP_001474562.1| DNA internalization-related competence protei...n ComEC/Rec2 [Shewanella sediminis HAW-EB3] gb|ABV37434.1| DNA internalization-related competence protein ComEC/Rec2 [Shewanella sediminis HAW-EB3] YP_001474562.1 1.7 33% ...

SarA is a negative regulator of Staphylococcus epidermidis biofilm formation

DEFF Research Database (Denmark)

Martin, Christer; Heinze, C.; Busch, M.

2012-01-01

Biofilm formation is essential for Staphylococcus epidermidis pathogenicity in implant-associated infections. Nonetheless, large proportions of invasive S. epidermidis isolates fail to show accumulative biofilm growth in vitro. We here tested the hypothesis that this apparent paradox is related...... virulence. Genetic analysis revealed that inactivation of sarA induced biofilm formation via over-expression of the giant 1 MDa extracellular matrix binding protein (Embp), serving as an intercellular adhesin. In addition to Embp, increased extracellular DNA (eDNA) release significantly contributed...... to biofilm formation in mutant 1585ΔsarA. Increased eDNA amounts indirectly resulted from up-regulation of metalloprotease SepA, leading to boosted processing of major autolysin AtlE, in turn inducing augmented autolysis and release of chromosomal DNA. Hence, this study identifies sarA as a negative...

NCBI nr-aa BLAST: CBRC-SARA-01-0756 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0756 gb|AAW70053.1| MRGX2 [Homo sapiens] gb|AAW70054.1| MRGX2 [Homo sapiens...] gb|AAW70055.1| MRGX2 [Homo sapiens] gb|AAW70070.1| MRGX2 [Homo sapiens] gb|AAW70083.1| MRGX2 [Homo sapiens] AAW70053.1 1e-66 52% ...

NCBI nr-aa BLAST: CBRC-SARA-01-1400 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1400 gb|AAW70053.1| MRGX2 [Homo sapiens] gb|AAW70054.1| MRGX2 [Homo sapiens...] gb|AAW70055.1| MRGX2 [Homo sapiens] gb|AAW70070.1| MRGX2 [Homo sapiens] gb|AAW70083.1| MRGX2 [Homo sapiens] AAW70053.1 4e-34 52% ...

NCBI nr-aa BLAST: CBRC-SARA-01-0472 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0472 sp|Q8NH41|OR4KF_HUMAN Olfactory receptor 4K15 dbj|BAC05798.1| sev...en transmembrane helix receptor [Homo sapiens] gb|EAW66492.1| olfactory receptor, family 4, subfamily K, member 15 [Homo sapiens] Q8NH41 1e-164 89% ...

NCBI nr-aa BLAST: CBRC-SARA-01-1206 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1206 sp|Q8NH41|OR4KF_HUMAN Olfactory receptor 4K15 dbj|BAC05798.1| sev...en transmembrane helix receptor [Homo sapiens] gb|EAW66492.1| olfactory receptor, family 4, subfamily K, member 15 [Homo sapiens] Q8NH41 1e-40 85% ...

Effect of Uncertainty Parameters in Blowdown and Reflood Models for OPR1000 LBLOCA Analysis

Energy Technology Data Exchange (ETDEWEB)

Huh, Byung Gil; Jin, Chang Yong; Seul, Kwangwon; Hwang, Taesuk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-05-15

KINS(Korea Institute of Nuclear Safety) has also performed the audit calculation with the KINS Realistic Evaluation Methodology(KINS-REM) to confirm the validity of licensee's calculation. In the BEPU method, it is very important to quantify the code and model uncertainty. It is referred in the following requirement: BE calculations in Regulatory Guide 1.157 - 'the code and models used are acceptable and applicable to the specific facility over the intended operating range and must quantify the uncertainty in the specific application'. In general, the uncertainty of model/code should be obtained through the data comparison with relevant integral- and separate-effect tests at different scales. However, it is not easy to determine these kinds of uncertainty because of the difficulty for evaluating accurately various experiments. Therefore, the expert judgment has been used in many cases even with the limitation that the uncertainty range of important parameters can be wide and inaccurate. In the KINS-REM, six heat transfer parameters in the blowdown phase have been used to consider the uncertainty of models. Recently, MARS-KS code was modified to consider the uncertainty of the five heat transfer parameters in the reflood phase. Accordingly, it is required that the uncertainty range for parameters of reflood models is determined and the effect of these ranges is evaluated. In this study, the large break LOCA (LBLOCA) analysis for OPR1000 was performed to identify the effect of uncertainty parameters in blowdown and reflood models.

Experimental study and modelling of pressure losses during reflooding of a debris beds

International Nuclear Information System (INIS)

Clavier, Remi

2015-01-01

This work deals with single and two-phase flow pressure losses in porous media. The aim is to improve understanding and modeling of momentum transfer inside particle beds, in relation with nuclear safety issues concerning the reflooding of debris beds during severe nuclear accidents. Indeed, the degradation of the core during such accidents can lead to the collapse of the fuel assemblies, and to the formation of a debris bed, which can be described as a hot porous medium. This thesis is included in a nuclear safety research project on coolability of debris beds during reflooding sequences. An experimental study of single and two-phase cold-flow pressure losses in particle beds is proposed. The geometrical characteristics of the debris and the hydrodynamic conditions are representative of the real case, in terms of granulometry, particle shapes, and flow velocities. The new data constitute an important contribution. In particular, they contain pressure losses and void fraction measurements in two-phase air-water flows with non-zero liquid Reynolds numbers, which did not exist before. Predictive models for pressure losses in single and two-phase flow through particle beds have been established from experimental data. Their structures are based on macroscopic equations obtained from the volume averaging of local conservation equations. Single-phase flow pressure losses can be described by a Darcy-Forchheimer law with a quadratic correction, in terms of filtration velocity, with a better-than-10 % precision. Numerical study of single-phase flows through porous media shows that this correlation is valid for disordered smooth particle beds. Two-phase flow pressure losses are described using a generalized Darcy-Forchheimer structure, involving inertial and cross flow terms. A new model is proposed and compared to the experimental data and to the usual models used in severe accident simulation codes. (author)

PLS models for determination of SARA analysis of Colombian vacuum residues and molecular distillation fractions using MIR-ATR

Directory of Open Access Journals (Sweden)

Jorge A. Orrego-Ruiz

2014-06-01

Full Text Available In this work, prediction models of Saturates, Aromatics, Resins and Asphaltenes fractions (SARA from thirty-seven vacuum residues of representative Colombian crudes and eighteen fractions of molecular distillation process were obtained. Mid-Infrared (MIR Attenuated Total Reflection (ATR spectroscopy in combination with partial least squares (PLS regression analysis was used to estimate accurately SARA analysis in these kind of samples. Calibration coefficients of prediction models were for saturates, aromatics, resins and asphaltenes fractions, 0.99, 0.96, 0.97 and 0.99, respectively. This methodology permits to control the molecular distillation process since small differences in chemical composition can be detected. Total time elapsed to give the SARA analysis per sample is 10 minutes.

NCBI nr-aa BLAST: CBRC-SARA-01-1746 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1746 ref|NP_653200.2| non-imprinted in Prader-Willi/Angelman syndrome ...1 [Homo sapiens] sp|Q7RTP0|NIPA1_HUMAN Non-imprinted in Prader-Willi/Angelman syndrome region protein 1 tpg|...DAA01477.1| TPA_exp: non-imprinted in Prader-Willi/Angelman syndrome 1 [Homo sapiens] NP_653200.2 1e-113 81% ...

The Remote Observatories of the Southeastern Association for Research in Astronomy (SARA)

Science.gov (United States)

Keel, William C.; Oswalt, Terry; Mack, Peter; Henson, Gary; Hillwig, Todd; Batcheldor, Daniel; Berrington, Robert; De Pree, Chris; Hartmann, Dieter; Leake, Martha; Licandro, Javier; Murphy, Brian; Webb, James; Wood, Matt A.

2017-01-01

We describe the remote facilities operated by the Southeastern Association for Research in Astronomy (SARA) , a consortium of colleges and universities in the US partnered with Lowell Observatory, the Chilean National Telescope Allocation Committee, and the Instituto de Astrofísica de Canarias. SARA observatories comprise a 0.96 m telescope at Kitt Peak, Arizona; one of 0.6 m aperture on Cerro Tololo, Chile; and the 1 m Jacobus Kapteyn Telescope at the Roque de los Muchachos, La Palma, Spain. All are operated using standard VNC or Radmin protocols communicating with on-site PCs. Remote operation offers considerable flexibility in scheduling, allowing long-term observational cadences difficult to achieve with classical observing at remote facilities, as well as obvious travel savings. Multiple observers at different locations can share a telescope for training, educational use, or collaborative research programs. Each telescope has a CCD system for optical imaging, using thermoelectric cooling to avoid the need for frequent local service, and a second CCD for offset guiding. The Arizona and Chile telescopes also have fiber-fed echelle spectrographs. Switching between imaging and spectroscopy is very rapid, so a night can easily accommodate mixed observing modes. We present some sample observational programs. For the benefit of other groups organizing similar consortia, we describe the operating structure and principles of SARA, as well as some lessons learned from almost 20 years of remote operations.

DETERMINACIÓN DE LAS FRACCIONES SARA DE ASFALTOS COLOMBIANOS ENVEJECIDOS AL MEDIO AMBIENTE EMPLEANDO CROMATOGRAFÍA LÍQUIDA EN COLUMNA DETERMINAÇÃO DAS FRAÇÕES SARA DE ASFALTOS COLOMBIANOS ENVELHECIDOS AO MÉDIO AMBIENTE EMPREGANDO CROMATOGRAFIA LÍQUIDA EM COLUNA DETERMINATION OF SARA FRACTIONS OF ENVIRONMENTALLY AGED COLOMBIAN ASPHALTS USING LIQUID CHROMATOGRAPHY COLUMN

Directory of Open Access Journals (Sweden)

Fredy Alberto Reyes

2012-06-01

Full Text Available En este artículo presentamos un método basado en cromatografía líquida en columna para cuantificar la composición química de los cementos asfálticos fabricados en Colombia, sometidos al medio ambiente, mediante la determinación de las fracciones SARA. El método fue aplicado sobre películas de asfalto 60/70 y 80/100 para determinar los cambios en la composición química del material luego de ser expuesto durante 12 meses a las condiciones de intemperie de la ciudad de Bogotá; los ensayos de SARA fueron efectuados para el asfalto original a 1, 3, 6, 9 y 12 meses respectivamente. Los fraccionamientos SARA evidenciaron que el envejecimiento produjo una disminución de la fracción de aromáticos y un incremento en la de asfaltenos respecto al asfalto no envejecido. La disminución de los compuestos aromáticos y de resinas pudo ser responsable del endurecimiento observado en los asfaltos, que presentaron una consistencia dura y quebradiza, lo que está de acuerdo con la obtención de índices coloidales elevados. El método empleado permitió establecer correlaciones entre la composición química del asfalto y sus propiedades mecánicas.Neste artigo apresentamos um método baseado em cromatografia líquida em coluna para quantificar a composição química dos cimentos asfálticos fabricados em Colômbia, submetidos ao médio ambiente, mediante a determinação das frações SARA. O método foi aplicado sobre películas de asfalto 60/70 e 80/100 para determinar as mudanças na composição química do material depois de ser exposto durante 12 meses às condições de intempérie da cidade de Bogotá; os ensaios de SARA foram efetuados para o asfalto original a 1, 3, 6, 9 e 12 meses respectivamente. Os fraccionamentos SARA evidenciaram que o envelhecimento produziu uma diminuição da fração de aromáticos e um incremento na de asfaltenos com respeito ao asfalto não envelhecido. A diminuição dos compostos aromáticos e de resinas p

Experimental and calculation results of the integral reflood test QUENCH-14 with M5 (registered) cladding tubes

International Nuclear Information System (INIS)

Stuckert, J.; Birchley, J.; Grosse, M.; Jaeckel, B.; Steinbrueck, M.

2010-01-01

The QUENCH-14 experiment investigated the effect of M5 (registered) cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (ISP-45) that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings. The PWR bundle configuration of QUENCH-14 with a single unheated rod, 20 heated rods, and four corner rods was otherwise identical to QUENCH-06. The test was conducted in principle with the same protocol as QUENCH-06, so that the effects of the change of cladding material could be observed more easily. Pre-test calculations were performed by the Paul Scherrer Institut (Switzerland) using the SCDAPSIM, SCDAP/RELAP5 and MELCOR codes. Follow-on post-test analyses were performed using SCDAP/RELAP5 and MELCOR as part of an ongoing programme of model validation and code assessment. Alternative oxidation correlations were used to examine the possible influence of the M5 (registered) cladding material on hydrogen generation, in comparison with Zircaloy-4. The experiment started with a pre-oxidation phase in steam, lasting ∼3000 s at ∼1500 K peak bundle temperature. After a further temperature increase to maximum bundle temperature of 2073 K the bundle was flooded with 2 g/s/rod water from the bottom. The peak temperature of ∼2300 K was measured on the bundle shroud, shortly after quench initiation. The electrical power was reduced to average value of 2 W/cm during the reflood phase to simulate effective decay heat level. Complete bundle cooling was reached in 300 s after reflood initiation. The development of the oxide layer growth during the test was essentially defined by measurements performed on the three Zircaloy-4 corner rods withdrawn successively from the bundle. The withdrawal of Zircaloy-4 and E110 corner rods after the test allowed a comparison of the different alloys in one test. One heated rod with M5 cladding was withdrawn after the test for a detailed analysis of oxidation degree and measurement of absorbed

Effect of a blockage length on the coolability during reflood in a 2 × 2 rod bundle with a 90% partially blocked region

Energy Technology Data Exchange (ETDEWEB)

Kim, Kihwan, E-mail: kihwankim@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Kim, Byung-Jae, E-mail: byoungjae@kaeri.re.kr [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseoung-Gu, Daejeon 34134 (Korea, Republic of); Choi, Hae-Seob, E-mail: hschoi@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Song, Chul-Hwa, E-mail: chsong@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of)

2017-02-15

Highlights: • This test was conducted to understand the effect of blockage length on the coolability. • Reflood tests were conducted with blockage simulators for various reflood rates. • The coolability in the downstream of the blockage region is significantly enhanced. - Abstract: If fuel rods are ballooned or rearranged during the reflood phase of a large break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the transient heat transfer behavior is entirely different with those of the intact fuel rods owing to the deformed blockage region. The coolability in the blocked region depends on a complex two-phase heat transfer with various thermal hydraulic conditions. In addition, the blockage characteristics, such as the blockage ratio, length, shape, and configurations, are also significant factors affecting the coolability. In the present study, reflood experiments were carried out to understand the effect of the blockage length upon the coolability by varying the reflooding rates. The experiments were performed in electrically heated 2 × 2 rod bundles with blockage simulators having the same blockage ratio but different blockage lengths. The characteristics of quenching and heat transfer were evaluated to investigate the influence of the blockage region on the coolability. The droplet behaviors were also observed by measuring the droplets velocity and size near the blockage region. The coolability in the downstream region of the blockage was significantly enhanced, owing to the reduced flow area of the sub-channel, intensification of turbulence, and the entrained droplets in the blockage region.

Numerical analysis of reflood simulation based on a mechanistic, best-estimate, approach by KREWET code

International Nuclear Information System (INIS)

Chun, Moon-Hyun; Jeong, Eun-Soo

1983-01-01

A new computer code entitled KREWET has been developed in an effort to improve the accuracy and applicability of the existing reflood heat transfer simulation computer code. Sample calculations for temperature histories and heat transfer coefficient are made using KREWET code and the results are compared with the predictions of REFLUX, QUEN1D, and the PWR-FLECHT data for various conditions. These show favourable agreement in terms of clad temperature versus time. For high flooding rates (5-15cm/sec) and high pressure (∼413 Kpa), reflood predictions are reasonably well predicted by KREWET code as well as with other codes. For low flooding rates (less than ∼4cm/sec) and low pressure (∼138Kpa), predictions show considerable error in evaluating the rewet position versus time. This observation is common to all the codes examined in the present work

Numerical analysis for reflood simulation based on a mechanistic, best-estimate, approach by KREWET code

International Nuclear Information System (INIS)

Chun, M.-H.; Jeong, E.-S.

1983-01-01

A new computer code entitled KREWET has been developed in an effort to improve the accuracy and applicability of the existing reflood heat transfer simulation computer code. Sample calculations for temperature histories and heat transfer coefficient are made using KREWET code and the results are compared with the predictions of REFLUX, QUENID, and the PWR-FLECHT data for various conditions. These show favorable agreement in terms of clad temperature versus time. For high flooding rates (5-15cm/sec) and high pressure (approx. =413 Kpa), reflood predictions are reasonably well predicted by KREWET code as well as with other codes. For low flooding rates (less than approx. =4cm/sec) and low pressure (approx. =138 Kpa), predictions show considerable error in evaluating the rewet position versus time. This observation is common to all the codes examined in the present work

Flow and heat transfer thermohydraulic modelisation during the reflooding phase of a P.W.R.'s core

International Nuclear Information System (INIS)

Raymond, Patrick

1978-04-01

Some generalities about L.O.C.A. are first recalled. The French experimental studies about Emergency Core Cooling System are briefly described. The different heat transfer mechanisms to take into account, according to the flow pattern in the dry zone, and the correlations or methods to calculate them, are defined. Then the Thermohydraulic code computer: FLIRA, which describe the reflooding phase, and a modelisation taking into account the different flow patterns are setted. A first interpretation of ERSEC experiments with a tubular test section shows that it is possible, with this modelisation and some classical heat transfer correlations, to describe the reflooding phase. [fr

Hemolytic disease of the fetus and newborn caused by an antibody to a low-prevalence antigen, anti-SARA.

Science.gov (United States)

Towns, Dale; Hannon, Judith; Hendry, Julia; Barnes, Janet; Goldman, Mindy

2011-09-01

The first case describing the SARAH (SARA) antigen occurred in 1990, in an Australian blood donor. Hemolytic disease of the fetus and newborn (HDFN) due to anti-SARA has not been previously described. We report a case of HDFN in a multiparous female. The pregnancy was unremarkable except that she was involved in a seemingly minor motor vehicle accident at 25 weeks' gestation. Routine prenatal antibody screening was negative throughout the pregnancy. She presented at 37 weeks' gestation because of decreased fetal movements. Labor was induced and a 2702-g infant male was delivered. The infant's hemoglobin was 49 g/L and the bilirubin was 153 µmol/L. Blood samples from the parents and infant were referred to Canadian Blood Services National Immunohematology Reference Laboratory and subsequently to the Australian Red Cross Red Cell Reference Service. The father's and infant's red blood cells were confirmed to be SARA positive, and the mother's plasma contained anti-SARA. The infant was successfully treated with a double-volume exchange transfusion. This is the first example of HDFN associated with this antibody. © 2011 American Association of Blood Banks.

Staphylococcus aureus Quorum Regulator SarA Targeted Compound, 2-[(Methylaminomethyl]phenol Inhibits Biofilm and Down-Regulates Virulence Genes

Directory of Open Access Journals (Sweden)

P. Balamurugan

2017-07-01

Full Text Available Staphylococcus aureus is a widely acknowledged Gram-positive pathogen for forming biofilm and virulence gene expressions by quorum sensing (QS, a cell to cell communication process. The quorum regulator SarA of S. aureus up-regulates the expression of many virulence factors including biofilm formation to mediate pathogenesis and evasion of the host immune system in the late phases of growth. Thus, inhibiting the production or blocking SarA protein might influence the down-regulation of biofilm and virulence factors. In this context, here we have synthesized 2-[(Methylaminomethyl]phenol, which was specifically targeted toward the quorum regulator SarA through in silico approach in our previous study. The molecule has been evaluated in vitro to validate its antibiofilm activity against clinical S. aureus strains. In addition, antivirulence properties of the inhibitor were confirmed with the observation of a significant reduction in the expression of representative virulence genes like fnbA, hla and hld that are governed under S. aureus QS. Interestingly, the SarA targeted inhibitor showed negligible antimicrobial activity and markedly reduced the minimum inhibitory concentration of conventional antibiotics when used in combination making it a more attractive lead for further clinical tests.

The Sun-Duffey mass effluents calculation model applied to bottom reflooding tests of a single tube performed at the CDTN

International Nuclear Information System (INIS)

Ladeira, L.C.D.; Rezende, H.C.

1993-01-01

A simple generalized model, developed by K.H. Sun and R.B. Duffey, is applied in this work to calculate the ratio of mass effluents during bottom reflooding of a single tube carried out at the CDTN/CNEN. The results of the benchmark experiments indicate that the accuracy on mass effluence ratio prediction can be within 15% by using the Sun-Duffey model. The reasonable agreement obtained between experimental data and model predictions suggest that it could be used for analysis of single tube reflood tests, in similar conditions. (author)

Upper plenum dump during reflood in PWR loss-of-coolant accident

International Nuclear Information System (INIS)

Sudo, Yukio; Griffith, Peter.

1981-01-01

Upper plenum dump during reflood in a large break loww-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood. The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnetude of water subcooling. (author)

Indicators of induced subacute ruminal acidosis (SARA) in Danish Holstein cows

DEFF Research Database (Denmark)

Danscher, Anne Mette; Li, Shucong; Andersen, Pia H.

2015-01-01

BACKGROUND: The prevalence of subacute ruminal acidosis (SARA) in dairy cows is high with large impact on economy and welfare. Its current field diagnosis is based on point ruminal pH measurements by oral probe or rumenocentesis. These techniques are invasive and inaccurate, and better markers fo...

Transferring generic SARA/OSHA training to US Department of Energy facilities

International Nuclear Information System (INIS)

White, A.; McKinley, T.

1989-01-01

The Technical Resources and Training Section staff at Oak Ridge National Laboratory have developed three extensive training programs for hazardous waste treatment, storage, and disposal facility workers a required by SARA/OSHA, 29 CFR 1910.120. The ORNL program is widely recognized as one of the best in the DOE system. ORNL and ORAU, who manages the Training Resources and Data Exchange (TRADE) network for DOE, entered into as cooperative relationship to respond to the many requests from DOE contractors for copies of the ORNL training materials. This discussion will describe the ORNL program and the process of turning it into a series of generic tools which can be used by additional DOE facilities to meet the training requirements established by SARA/OSHA, 20 CFR 1910.120. The speakers will describe how the materials are being used by DOE facilities as well as plans for additional resources to be developed through TRADE. 5 refs

NCBI nr-aa BLAST: CBRC-SARA-01-1597 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1597 sp|Q9XT58|ADRB3_SHEEP Beta-3 adrenergic receptor (Beta-3 adrenoce...ptor) (Beta-3 adrenoreceptor) gb|AAG31165.1|AF314202_1 beta 3 adrenergic receptor [Ovis aries] gb|AAG31167.1|AF314204_1 beta 3 adrene...rgic receptor [Ovis aries] gb|ABB71185.1| beta 3 adrenergic reecptor [Ovis aries] Q9XT58 1e-140 75% ...

SARAS 2 Constraints on Global 21 cm Signals from the Epoch of Reionization

Science.gov (United States)

Singh, Saurabh; Subrahmanyan, Ravi; Udaya Shankar, N.; Sathyanarayana Rao, Mayuri; Fialkov, Anastasia; Cohen, Aviad; Barkana, Rennan; Girish, B. S.; Raghunathan, A.; Somashekar, R.; Srivani, K. S.

2018-05-01

Spectral distortions in the cosmic microwave background over the 40–200 MHz band are imprinted by neutral hydrogen in the intergalactic medium prior to the end of reionization. This signal, produced in the redshift range z = 6–34 at the rest-frame wavelength of 21 cm, has not been detected yet; and a poor understanding of high-redshift astrophysics results in a large uncertainty in the expected spectrum. The SARAS 2 radiometer was purposely designed to detect the sky-averaged 21 cm signal. The instrument, deployed at the Timbaktu Collective (Southern India) in 2017 April–June, collected 63 hr of science data, which were examined for the presence of the cosmological 21 cm signal. In our previous work, the first-light data from the SARAS 2 radiometer were analyzed with Bayesian likelihood-ratio tests using 264 plausible astrophysical scenarios. In this paper we reexamine the data using an improved analysis based on the frequentist approach and forward-modeling. We show that SARAS 2 data reject 20 models, out of which 15 are rejected at a significance >5σ. All the rejected models share the scenario of inefficient heating of the primordial gas by the first population of X-ray sources, along with rapid reionization. Joint Astronomy Program, Indian Institute of Science, Bangalore 560012, India.

BWR Refill-Reflood Program. Final report

International Nuclear Information System (INIS)

Myers, L.L.

1983-09-01

The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

Fish fauna recovery in a newly re-flooded Mediterranean coastal lagoon

Science.gov (United States)

Koutrakis, Emmanuil; Sylaios, Georgios; Kamidis, Nikolaos; Markou, Dimitrios; Sapounidis, Argyris

2009-08-01

Drana Lagoon, located at the NW site of Evros River Delta, was drained in 1987 and re-flooded in 2004 within the framework of an integrated wetland restoration project. This study presents the results of a monitoring program of the lagoon's oceanographic, water quality and fish fauna characteristics, during the pre- and post-restoration period. Results depict the presence of high salinity water (up to 41) due to seawater intrusion, strong evaporation in its interior and inadequate freshwater inflows. Overall, nutrient levels were low depicting local changes. Tidal variability at the mouth was approximately 0.2 m, producing high velocity tidal currents (up to 0.75 m/s). Eleven fish fauna species were collected; seven species were caught in both the inlet channel and the lagoon during the pre-restoration period and nine species in the post-restoration period. Atherina boyeri (37.6%) and Pomatoschistus marmoratus (31.7%) dominated the lagoon during the post-restoration period. Most of the A. boyeri specimens (88.5%) were caught inside the lagoon, while P. marmoratus had an almost equal distribution in the inlet channel and the lagoon (56.3% and 43.7% respectively). The presence of species of the Mugilidae family (5.2% total average catches after lagoon re-flooding) was mainly in the inlet channel (12.6% of the average catches) and not inside the lagoon (only 1.3% of the average catches). The small number of fish species inhabiting the lagoon might be the result of the recent restoration or it could be related with the increased water flow observed at the lagoon mouth during the flood and ebb tidal phases, and also in the presence of a smooth bank in the concrete waterspout that connects the entrance channel with the lagoon. The limited presence of the Mugilidae juveniles inside the lagoon could be related to the prevailing tidal inlet dynamics (i.e. strong ebb flow at lagoon inlet), thus preventing the species to enter the lagoon. In order to restore the lagoon

NCBI nr-aa BLAST: CBRC-SARA-01-0195 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0195 ref|NP_036916.1| cannabinoid receptor 1 (brain) [Rattus norvegicu...s] sp|P20272|CNR1_RAT Cannabinoid receptor 1 (CB1) (CB-R) (Brain-type cannabinoid receptor) emb|CAA39332.1| CB1 cann...abinoid receptor [Rattus norvegicus] gb|AAA99067.1| neuronal cannabinoid receptor gb|EDL98589.1| cann...abinoid receptor 1 (brain) [Rattus norvegicus] prf||1613453A cannabinoid receptor NP_036916.1 0.0 97% ...

Effect of heat transfer in the fog region during core reflooding

International Nuclear Information System (INIS)

Rouai, N. M.; El-sawy, H. M.

1993-01-01

Core reflooding following a loss of coolant accident (LOCA) in a pressurized water reactor (PWR) received considerable attention during the past thirty years. In this paper a one dimensional model is used to study the effect of the heat transfer in the fog region ahead of the wet front reflooding rate of a cylindrical fuel element following a LOCA in a PWR. The heat conduction equation in the cladding is solved in coordinate system moving with the wet front under a variety of condition to investigate the effects of such parameters as the initial cladding surface temperature, the decay heat generation rate in the fuel and the mode of heat transfer prevailing. The cladding surface is divided into three axial regions according to the mechanism of heat transfer, namely, a boiling region behind the wet front, a fog region ahead of the wet front and a dry region further downstream of the wet front. The effect of changing the heat transfer coefficient in the fog region on the rewetting rate and on the fog length is investigated. The results of this simple model show that increasing the heat transfer in the fog region increases the rewetting velocity and consequently decreases the fog length. The results are in general agreement with a more accurate two-dimensional model and experimental data. (author)

Safety Research Program for Light Water Reactors. Technical report 2: BMFT support project RS 0036 B. Reflooding experiments with regard to primary circuits (PKL) instrumentation of experimental setup

International Nuclear Information System (INIS)

Schweickert, H.; Mandl, R.

The reflooding of the hot core of a PWR will be investigated in a model of the complete primary system. The demands that the instrumentation must meet as well as a description of the measurement methods used in the circuit are described. Data on the efficiency of the instruments, error estimates and constructive solutions to design problems are also given

Cold leg injection reflood test results in the SCTF Core-I under constant system pressure

International Nuclear Information System (INIS)

Adachi, Hiromichi; Iwamura, Takamichi; Sobajima, Makoto; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio.

1990-08-01

The Slab Core Test Facility (SCTF) was constructed to investigate two-dimensional thermal-hydrodynamics in the core and the interaction in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a postulated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report describes the analytical results on the system behavior observed in the SCTF Core-I cold leg injection tests, S1-14 (Run 520), S1-15 (521), S1-16 (522), S1-17 (523), S1-20 (530), S1-21 (531), S1-23 (536) and S1-24 (537), performed under constant system pressure condition during transient. Major discussion items are: (1) steam binding, (2) U-tube oscillations, (3) bypass of ECC water (4) core cooling behavior, (5) effect of vent valve and (6) other parameter effects. These results give us very useful information and suggestion on reflood behavior. (author)

NCBI nr-aa BLAST: CBRC-SARA-01-1750 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1750 gb|ABR53744.1| opsin 1 short-wavelength senstive protein [Daubentonia madagascar...iensis] gb|ABR53745.1| opsin 1 short-wavelength senstive protein [Daubentonia madagascariensi...s] gb|ABR53746.1| opsin 1 short-wavelength senstive protein [Daubentonia madagascariensis] gb|ABR53747.1| op...sin 1 short-wavelength senstive protein [Daubentonia madagascariensis] gb|ABR5374...8.1| opsin 1 short-wavelength senstive protein [Daubentonia madagascariensis] gb|ABR53749.1| opsin 1 short-w

A model for dispersed flow heat transfer in rod bundles during reflood

International Nuclear Information System (INIS)

Wong, S.

1980-01-01

The present model calculates the heat transfer characteristics of the non-equilibrium dispersed droplet flow regime above the quench front during reflood by solving simultaneously the wall-to-vapor interactions, wall-to-droplet interactions and vapor-to-droplet interactions by an iterative numerical method. The unique feature in the present study is various heat transfer mechanisms are combined in an overall energy balance equation, and the convective heat transfer to vapor is obtained by calculating the vapor temperature distributions at the heated walls. The reactor rod bundle geometry, axial variations of vapor temperature and flow properties, radiative heat transfers, and enhancement of heat transfer due to turbulence are considered carefully, so that the present model could be used to predict PWR (Pressurized Water Reactor) reflood heat transfers, and hence the fuel cladding wall temperature transients. In order to achieve closure of the problem formulations, the droplet size and its motion are determined from the FLECHT (Full Length Emergency Cooling Heat Transfer Program) low flooding rate series consine axial power shape test data. The model is then verified by comparing the heat transfer predictions with FLECHT low flooding rate series skewed axial power shape test data. Comparisons of predictions with data show satisfactory agreements

COBRA/TRAC analysis of two-dimensional thermal-hydraulic behavior in SCTF reflood tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Ohnuki, Akira; Sobajima, Makoto; Adachi, Hiromichi

1987-01-01

The effects of radial power distribution and non-uniform upper plenum water accumulation on thermal-hydraulic behavior in the core were observed in the reflood tests with Slab Core Test Facility (SCTF). In order to examine the predictability of these two effects by a multi-dimensional analysis code, the COBRA/TRAC calculations were made. The calculated results indicated that the heat transfer enhancement in high power bundles above quench front was caused by high vapor flow rate in those bundles due to the radial power distribution. On the other hand, the heat transfer degradation in the peripheral bundles under the condition of non-uniform upper plenum water accumulation was caused by the lower flow rates of vapor and entrained liquid above the quench front in those bundles by the reason that vapor concentrated in the center bundles due to the cross flow induced by the horizontal pressure gradient in the core. The above-mentioned two-dimensional heat transfer behaviors calculated with the COBRA/TRAC code is similar to those observed in SCTF tests and therefore those calculations are useful to investigate the mechanism of the two-dimensional effects in SCTF reflood tests. (author)

Changes in Microbiota in Rumen Digesta and Feces Due to a Grain-Based Subacute Ruminal Acidosis (SARA) Challenge

DEFF Research Database (Denmark)

Plaizier, Jan C.; Li, Shucong; Danscher, Anne Mette

2017-01-01

The effects of a grain-based subacute ruminal acidosis (SARA) challenge on bacteria in the rumen and feces of lactating dairy cows were determined. Six lactating, rumen-cannulated Danish Holstein cows were used in a cross-over study with two periods. Periods included two cows on a control diet...... and two cows on a SARA challenge. The control diet was a total mixed ration containing 45.5% dry matter (DM), 43.8% DM neutral detergent fiber, and 19.6% DM starch. The SARA challenge was conducted by gradually substituting the control diet with pellets containing 50% wheat and 50% barley over 3 days...... to reach a diet containing 55.6% DM, 31.3% DM neutral detergent fiber, and 31.8% DM starch, which was fed for four more days. Rumen fluid samples were collected at day 7 and 10 of experimental periods. Feces samples were collected on days 8 and 10 of these periods. Extracted DNA from the rumen and feces...

Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

International Nuclear Information System (INIS)

Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

1995-01-01

Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

BERTHA: a programme for the thermal/hydraulic analysis of reflooding experiments

International Nuclear Information System (INIS)

Pearson, K.G.; Cooper, C.A.

1985-04-01

In the event of a large break loss-of coolant accident in a PWR the normal cooling would be restored by reflooding the dry overheated reactor core from below. A model, BERTHA, of heat transfer in this dry region is presented. It includes a film boiling and dispersed flow region and explicitly represents the effect of spacer grids. In parallel channel mode it can calculate the effect of partial flow blockage. Predictions of the model are compared with experimental data and show good agreement in both blocked and unblocked configurations. (U.K.)

Sara John: Ethnisierte Arbeit. Eine feministische Perspektive. Marburg: Tectum Wissenschaftsverlag 2009.

Directory of Open Access Journals (Sweden)

Grit Grigoleit

2011-03-01

Full Text Available Auch bei steigender Erwerbsbeteiligung von Frauen ist der deutsche Arbeitsmarkt von einer weitverbreitenden Chancenungleichheit gekennzeichnet. Die Lebenswirklichkeiten von Migrant/-innen und ihre Einbindung in die vergeschlechtlichten Prozesse am Arbeitsmarkt wurden bislang nicht systematisch erfasst. An diesem Punkt setzt Sara John an, indem sie theoretische Konzeptionen zur Vergeschlechtlichung und zur Ethnisierung auf dem Arbeitsmarkt zusammenführt. In einem multidisziplinären Ansatz werden die zahlreichen Verschränkungen um das Phänomen ‚ethnisierte Arbeit‘ aufgegriffen, die vor dem Hintergrund der Debatte um Deutschland als Einwanderungsland zunehmend an Bedeutung und Brisanz gewinnen.Although women’s labor force participation is rising, the German job market is characterized by a widespread lack of equal opportunities. Thus far, the everyday realities of migrants and their integration into the gendered processes on the job market have not been collected systematically. This is where Sara John begins her study by combining theoretical conceptions about gendering and ethnicizing on the job market. In a multidisciplinary approach, several entanglements surrounding the phenomenon ‘ethnicized labor’ are taken into account. These entanglements keep gaining importance and topicality against the backdrop of Germany as a country of immigration.

Layering of life (Sara novel of Peter Sarić)

OpenAIRE

Kostić, Dragomir J.

2015-01-01

Novel Sara of Petar Sarić consists of two parts; in it are processed or present two wars, two major wars in the region of Montenegro and Herzegovina, the First and Second World War. However, it is more novel about divisions within the family and the man himself, (and infamous assault of godfather Luka on Sarah also and his murder are in that function), in the first part; and on the divisions among the people, in general, in the second part of the novel. The second part is, in fact, the image ...

NCBI nr-aa BLAST: CBRC-SARA-01-1608 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1608 ref|NP_001832.1| cannabinoid receptor 2 (macrophage) [Homo sapien...s] sp|P34972|CNR2_HUMAN Cannabinoid receptor 2 (CB2) (CB-2) (CX5) emb|CAA52376.1| CB2 (peripheral) cannabino...id receptor [Homo sapiens] emb|CAD22548.1| peripheral cannabinoid receptor CB2 [Homo sapiens] emb|CAD22549.1| peripheral cann...abinoid receptor CB2 [Homo sapiens] gb|AAO92299.1| cannabinoid r...eceptor 2 [Homo sapiens] emb|CAI14799.1| cannabinoid receptor 2 (macrophage) [Homo sapiens] emb|CAJ42137.1| cann

Adaptation of Toodee-2 computer code for reflood analysis in Angra-1 reactor

International Nuclear Information System (INIS)

Praes, J.G.L.; Onusic Junior, J.

1981-01-01

A method of calculation the heat transfer coefficient used in Toodee-2 computer code for core reflood analysis in a loss of coolant accident, is presented. Preliminary results are presented with the use of heat transfer correlations based on FLECHT experiments adequate to a geometric arrangement such as 16 x 16 (Angra I). Optional calculations are suggested for the heat transfer coefficients when the cooling of fuel cladding by steam is used. (Author) [pt

Saras Cranes in Palwal District in Southern Haryana are Asking for Immediate Attention for Their Last Rescue Effort

Directory of Open Access Journals (Sweden)

Tirshem Kumar Kaushik

2015-05-01

Full Text Available Saras Cranes Grus antigone are endangered birds of open wetlands with highly worrying depletion trends being witnessed related with disappearance of marshy and shallow perennial, expansive wetlands throughout northern India. Alongside, massive hunting in 18th, 19th and 20th centuries and even today is another serious cause for their worrisome deterioration. Also, destruction of nests, eggs, fledglings and adults by aboriginals indeliberately or deliberately is causing these cranes to perish sooner than latter, completely. Now, Saras Cranes are found in limited number and domain as four populations in the entire world including India, China, Burma, South East Asia and northern Australia. The population of Indian Saras Crane is pitiably restricted to Etawa and Mainpuri districts of Uttar Pradesh. Stray birds of this species are restricted to Kanha National Park in Madhya Pradesh and in some parts of Gujarat and Assam. It is interesting to note that few pairs have been seen in Faridabad and Palwal districts in southern Haryana, India. These need to be protected and conserved.

Sara Wasson and Emily Alder, Gothic Science Fiction 1980-2010

OpenAIRE

Beaulé, Sophie

2014-01-01

With their collection of essays Gothic Science Fiction 1980-2010, Sara Wasson and Emily Alder illustrate the richness of gothic tropes in contemporary forms, from novels and movies to card games. More than cliché, melodrama, or gore, the “gothick” (to borrow Adam Roberts’s term, xi) allows for the hybridity in contemporary production, especially in science fiction, that the collected articles examine. The book is divided into three parts, “Redefining Genres”, “Biopower and Capital”, and “Gend...

The Workplace Literacy System Project (WLS). Final Performance Report.

Science.gov (United States)

Poulton, Bruce R.

The Workplace Literacy System Project (WLS) prepared interactive CD-ROM discs containing about 50 hours of instruction and drill in basic skills presented within the context of the textile/apparel manufacturing industry. The project was conducted at a Sara Lee knit products plant in North Carolina. During the project, literacy task analyses were…

Analysis of the reflood experiment by RELAP5/MOD2 code

International Nuclear Information System (INIS)

Prosek, A.; Stritar, A.

1990-01-01

The analysis of the reflood experiment on the test rig Achilles has been performed. The analysis has been done by the RELAP5/MOD2 code after the results of the experiment had been released. The experiment has been analyze in several other laboratories around the world. Our results are comparable to other analyses and are in the range of RELAP5/MOD2 capabilities. Two analyses have been done: the core only and the complete system. Computed clad temperatures in the first case are higher than measured, in the second case they are somewhat lower. (author)

Women Empowerment in the Realms of Institutionalized Religion and Patriarchy: El Saadawi’s Firdaus and Yezierska’s Sara as Examples

Directory of Open Access Journals (Sweden)

Abdullah K. Shehabat

2016-09-01

Full Text Available This paper explains how the two protagonists, Firdaus and Sara, successfully paved their own ways in search of self-liberation despite the authoritarian patriarchy and institutionalized religions that plagued them. El Saadawi's Woman at Point Zero and Yezierska’s Bread Givers represent the fruitful struggle these protagonists experienced as they come to forge an identity and be themselves. The paper argues that the protagonists manage to free themselves, establish their own spiritual homes at their own homes and assert the potentials of their femininity despite their endings. Empowered by the powers of reading, strong will and meticulous work, the protagonists were able to realize their own material independence and achieve their lifelong ambitions. However, through Firdaus' and Sara's journeys of breaking their silence, they were subject to different patterns of self-annihilation. While Firdaus was sentenced to death for killing a pimp, Sara embraced living under the hegemony of an authoritarian husband.

The new species of Mysidacea (Crustacea), Anchialina labatus and Gastrosaccus sarae, from south west Australia

Digital Repository Service at National Institute of Oceanography (India)

Panampunnayil, S.U.

on the third segment of the mandibular palp and by the modification of the exopod of the third pleopod of the male. G. sarae is distinguished from the other species by the shape and armature of the telson...

A simple blowdown code for SUPER-SARA loop conditions

International Nuclear Information System (INIS)

Fritz, G.

1981-01-01

The Super Sara test programme (SSTP) is aimed to study in pile the fuel and cluster behaviour under two types of accident conditions: - the ''Large break loss of coolant'' condition (LB-Loca), - the ''Severe fuel damage'' (SFD) in a boildown caused by a small break. BIVOL was developed for the LB-Loca situation. This code is made for a loop where essentially two volumes define the thermohydraulics during the blowdown. In the SUPERSARA loop these two volumes are represented by the hot leg and cold leg pipings together with the respective upper and lower plenum of the test section

NCBI nr-aa BLAST: CBRC-SARA-01-1691 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1691 ref|NP_000854.1| 5-hydroxytryptamine (serotonin) receptor 1B [Hom...o sapiens] ref|NP_001009102.1| 5-hydroxytryptamine (serotonin) receptor 1B [Pan troglodytes] sp|P28222|5HT1B...HT-1B) (Serotonin receptor 1B) (5-HT1B) gb|AAA58675.1| serotonin 1Db receptor gb|AAA36029.1| serotonin recep...tor gb|AAA36030.1| 5-hyroxytryptamine 1D receptor dbj|BAA01763.1| serotonin 1B receptor [Homo sapiens] gb|AAA60316.1| serotonin... 1D receptor emb|CAB51537.1| 5-hydroxytryptamine (serotonin) r

NCBI nr-aa BLAST: CBRC-SARA-01-1942 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1942 ref|NP_000669.1| alpha-1D-adrenergic receptor [Homo sapiens] sp|P...25100|ADA1D_HUMAN Alpha-1D adrenergic receptor (Alpha 1D-adrenoceptor) (Alpha 1D-adrenoreceptor) (Alpha-1A adrenergic... receptor) (Alpha adrenergic receptor 1a) gb|AAB60351.1| adrenergic alpha-1a receptor protein gb|AAB59487.1| alpha 1a/d adre...nergic receptor dbj|BAA06222.1| alpha1A/D adrenergic rec...eptor [Homo sapiens] emb|CAH70478.1| adrenergic, alpha-1D-, receptor [Homo sapiens] emb|CAC00601.2| adrenergic

Reflooding Experiment on BETA Test Loop: The Effects of Inlet Temperature on the Rewetting Velocity

International Nuclear Information System (INIS)

Khairul H; Anhar R Antariksawan; Edy Sumarno; Kiswanta; Giarno; Joko P; Ismu Handoyo

2003-01-01

Loss of Coolant Accident (LOCA) on Nuclear Reactor Plant is an important topic because this condition is a severe accident that can be postulated. The phenomenon of LOCA on Pressurized Water Reactor (PWR) can be divided in three stages, e.g.: blowdown, refill and reflood. In the view of Emergency Coolant System evaluation, the reflood is the most important stage. In this stage, an injection of emergency water coolant must be done in a way that the core can be flooded and the overheating can be avoid. The experiment of rewetting on BETA Test Loop had been conducted. The experiment using one heated rod of the test section to study effects of inlet temperature on the wetting velocity. Results of the series of experiments on 2,5 lt/min flow rate and variable of temperature : 28 o C, 38 o C, 50 o C, 58 o C it was noticed that for 58 o C inlet temperature of test section and 572 o C rod temperature the rewetting phenomenon has been observed. The time of refill was 32.81 sec and time of rewetting was 42.87 sec. (author)

REFLA-1D/MODE3: a computer code for reflood thermo-hydrodynamic analysis during PWR-LOCA

International Nuclear Information System (INIS)

Murao, Yoshio; Okubo, Tsutomu; Sugimoto, Jun; Iguchi, Tadashi; Sudoh, Takashi.

1985-02-01

This manual describes the REFLA-1D/MODE3 reflood system analysis code. This code can solve the core thermo-hydrodynamics under forced flooding conditions and gravity feed conditions in a system similar to FLECHT-SET Phase A. This manual describes the REFLA-1D/MODE3 models and provides application information required to utilize the code. (author)

Nuclear spectroscopy of exotic nuclei with the SARA/IGISOL facility

International Nuclear Information System (INIS)

Beraud, R.; Emsallem, A.; Astier, A.; Duffait, R.; Aerje, J.; Aeystoe, J.; Jauho, P.; Barneoud, D.; Genevey, J.; Gizon, A.

1995-01-01

Some recent decay studies of neutron-rich and proton-rich nuclei are presented for nuclear structure investigations far off the valley of stability. The experiments, carried out at SARA, are based either on charged particle-induced fission of 238 U or on HI-induced fusion-evaporation reactions in combination with the IGISOL technique. The basic principle of this latter is recalled together with its advantages and limitations. The spectroscopic results obtained in three different regions of the chart of nuclei are sketched. (authors). 30 refs., 7 figs

A study on quench phenomena during reflood phase, 1

International Nuclear Information System (INIS)

Murao, Yoshio; Sudoh, Takashi

1977-03-01

Based on the observation with an outside-heated quartz tube experiment of the reflood phase, three quench modes for bottom flooding are proposed : 1) liquid column type, 2) dryout type, 3) droplet-rewetting type. Using Blair's correlation for quench velocity, the approximate correlation for maximum liquid superheat, the assumption that the heat transfer upstream of the quench front is a function of the local liquid subcooling and the data of PWR-FLECHT experiments, the correlation for quench velocity of the liquid column type and of the dryout type are obtained. The quench temperature for the droplet-rewetting type is also derived. These relations are compared with the results of PWR-FLECHT Group 1 experiments and of Piggott and Porthouse's experiments. The agreements among them are fairly good. (auth.)

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), Version 5.0. Volume 5, Systems Analysis and Risk Assessment (SARA) tutorial manual

International Nuclear Information System (INIS)

Sattison, M.B.; Russell, K.D.; Skinner, N.L.

1994-07-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs) primarily for nuclear power plants. This volume is the tutorial manual for the Systems Analysis and Risk Assessment (SARA) System Version 5.0, a microcomputer-based system used to analyze the safety issues of a open-quotes familyclose quotes [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. A series of lessons is provided that guides the user through some basic steps common to most analyses performed with SARA. The example problems presented in the lessons build on one another, and in combination, lead the user through all aspects of SARA sensitivity analysis capabilities

The technical status of the supersara project at its termination

International Nuclear Information System (INIS)

Markovina, A.; Randles, J.

1987-01-01

On March 10th 1983 the Council of Ministers of the European Community adopted the decision to abandon the SuperSARA project. In the frame of this decision, the Management of the SuperSARA project formulated the guidelines to close the project, which included the preparation of the present final report. This report presents the final status of the project, which means that it reflects the situation in March 1983, with a few updates introduced shortly afterwards in conformity with the safety requirements which were being revised at that time. The aim of this report is to give a general description of activities which were carried on to implement the experimental programme and to illustrate some important achievements reached in specific areas. In addition, it is also intended to provide the access key to the documentation which was produced in the past years, so as to make the most significant technical material easily available for any possible future use

Evaluation report on CCTF Core-II reflood test second shakedown test, C2-SH2 (Run 54)

International Nuclear Information System (INIS)

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

1985-03-01

A low power test (the initial averaged linear power density = 1.18 kW/m) and the base case test (1.4 kW/m) were performed with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute, in order to study the effect of the power on the reflood phenomena. The former linear power density corresponds nearly to the scaled linear power density based on the current safety evaluation criterio. During the early period of the reflood ( 200s) the heat transfer coefficient became higher and resultantly the quench front advanced faster in the low power test. The core flooding rate was nearly identical between both tests, independently of the different power. The insensitiveness of the power to the core flooding rate was also observed in FLECHT-SET performed in the USA. A significatn large differential pressure oscillation at ECC ports was experienced in the low power test, and it may be important for the long term core cooling although it has not been taken note on the previous studies. (author)

NCBI nr-aa BLAST: CBRC-SARA-01-0948 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-0948 ref|NP_473371.1| MAS-related GPR, member X2 [Homo sapiens] sp|Q96...LB1|MRGX2_HUMAN Mas-related G-protein coupled receptor member X2 gb|AAK91805.1| G protein-coupled receptor [Homo sapiens...] dbj|BAB89339.1| putative G-protein coupled receptor [Homo sapiens] dbj|BAC06030.1| seven trans...membrane helix receptor [Homo sapiens] gb|AAH63450.1| MAS-related GPR, member X2 [Homo sapiens...] gb|AAW70056.1| MRGX2 [Homo sapiens] gb|AAW70057.1| MRGX2 [Homo sapiens] gb|AAW70058.1| MRGX2 [Homo sapiens

NCBI nr-aa BLAST: CBRC-SARA-01-1066 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1066 ref|NP_473371.1| MAS-related GPR, member X2 [Homo sapiens] sp|Q96...LB1|MRGX2_HUMAN Mas-related G-protein coupled receptor member X2 gb|AAK91805.1| G protein-coupled receptor [Homo sapiens...] dbj|BAB89339.1| putative G-protein coupled receptor [Homo sapiens] dbj|BAC06030.1| seven trans...membrane helix receptor [Homo sapiens] gb|AAH63450.1| MAS-related GPR, member X2 [Homo sapiens...] gb|AAW70056.1| MRGX2 [Homo sapiens] gb|AAW70057.1| MRGX2 [Homo sapiens] gb|AAW70058.1| MRGX2 [Homo sapiens

NCBI nr-aa BLAST: CBRC-SARA-01-1746 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1746 ref|NP_705806.1| non-imprinted in Prader-Willi/Angelman syndrome ...1 [Mus musculus] sp|Q8BHK1|NIPA1_MOUSE Non-imprinted in Prader-Willi/Angelman syndrome region protein 1 homo...protein product [Mus musculus] gb|AAH55828.1| Non imprinted in Prader-Willi/Angel...man syndrome 1 homolog (human) [Mus musculus] gb|EDL21870.1| non imprinted in Prader-Willi/Angelman syndrome 1 homolog (human) [Mus musculus] NP_705806.1 1e-113 81% ... ...log gb|AAM34534.1| non-imprinted in Prader-Willi/Angelman syndrome 1 [Mus musculus] dbj|BAC32809.1| unnamed

Reading, Laterality, and the Brain: Early Contributions on Reading Disabilities by Sara S. Sparrow

Science.gov (United States)

Fletcher, Jack M.; Morris, Robin D.

2014-01-01

Although best known for work with children and adults with intellectual disabilities and autism spectrum disorders, training in speech pathology and a doctorate in clinical psychology and neuropsychology was the foundation for Sara Sparrow's long-term interest in reading disabilities. Her first papers were on dyslexia and laterality, and the…

The LP-FP-2 severe fuel damage scenario and discussion of the relative influence of the transient and reflood phases in affecting the final condition of the bundle

International Nuclear Information System (INIS)

Modro, S.M.; Carboneau, M.L.

1990-01-01

The purpose of this paper is to review the evidence from the OECD LP-FP-2 experiment that a high temperature excursion occurred within the center fuel module (CFM) during the reflood portion of the test, was caused by rapid metal-water reaction. It is shown that this reflood scenario explains many perplexing observations from the experiment, in particular, the small amount of fission products and hydrogen transported to the blowdown suppression tank (BST) as compared with the larger quantities trapped within the primary coolant system (PCS). The timing and destruction of the CFM upper tie plate, as well as the transport of fuel debris to the top of this plate, are also explained. In general, all measurements, observations, and analyses of the LP-FP-2 data indicate that most of the CFM damage occurred during a relatively short period of time coincident with the reflood portion of the experiment. 4 refs., 6 figs

Social humanoid robot SARA: development of the wrist mechanism

Science.gov (United States)

Penčić, M.; Rackov, M.; Čavić, M.; Kiss, I.; Cioată, V. G.

2018-01-01

This paper presents the development of a wrist mechanism for humanoid robots. The research was conducted within the project which develops social humanoid robot Sara - a mobile anthropomorphic platform for researching the social behaviour of robots. There are two basic ways for the realization of humanoid wrist. The first one is based on biologically inspired structures that have variable stiffness, and the second one on low backlash mechanisms that have high stiffness. Our solution is low backlash differential mechanism that requires small actuators. Based on the kinematic-dynamic requirements, a dynamic model of the robot wrist is formed. A dynamic simulation for several hand positions was performed and the driving torques of the wrist mechanism were determined. The realized wrist has 2 DOFs and enables movements in the direction of flexion/extension 115°, ulnar/radial deviation ±45° and the combination of these two movements. It consists of a differential mechanism with three spur bevel gears, two of which are driving and identical, while the last one is the driven gear to which the robot hand is attached. Power transmission and motion from the actuator to the input links of the differential mechanism is realized with two parallel placed identical gear mechanisms. The wrist mechanism has high carrying capacity and reliability, high efficiency, a compact design and low backlash that provides high positioning accuracy and repeatability of movements, which is essential for motion control.

EFLOD code for reflood heat transfer

International Nuclear Information System (INIS)

Gay, R.R.

1979-01-01

A computer code called EFLOD has been developed for simulation of the heat transfer and hydrodynamics of a nuclear power reactor during the reflood phase of a loss-of-coolant accident. EFLOD models the downcomer, lower plenum, core, and upper plenum of a nuclear reactor vessel using seven control volumes assuming either homogeneous or unequal-velocity, unequal-temperature (UVUT) models of two-phase flow, depending on location within the vessel. The moving control volume concept in which a single control volume models the quench region in the core and moves with the core liquid level was developed and implemented in EFLOD so that three control volumes suffice to model the core region. A simplified UVUT model that assumes saturated liquid above the quench front was developed to handle the nonhomogeneous flow situation above the quench region. An explicit finite difference routine is used to model conduction heat transfer in the fuel, gap, and cladding regions of the fuel rod. In simulation of a selected FLECHT-SET experimental run, EFLOD successfully predicted the midplane maximum temperature and turnaround time as well as the time-dependent advance of the core liquid level. However, the rate of advancement of the quench level and the ensuing liquid entrainment were overpredicted during the early part of the transient

Results of the QUENCH-12 experiment on reflood of a VVER-type bundle

Energy Technology Data Exchange (ETDEWEB)

Stuckert, J.; Grosse, M.; Heck, M.; Schanz, G.; Sepold, L.; Stegmaier, U.; Steinbrueck, M. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Materialforschung, Programm Nukleare Sicherheitsforschung; Goryachev, A.; Ivanova, I. [RIAR (FSUE SSC-RIAR) Dimitrovgrad (Russian Federation)

2008-09-15

The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor. The QUENCH test bundle with a total length of approximately 2.5 m usually consists of 21 fuel rod simulators of Western PWR (Pressurized Water Reactor) geometry. The QUENCH-12 test bundle, however, which was set up to investigate the effects of VVER materials and bundle geometry (hexagonal lattice) on core reflood consisted of 31 fuel rod simulators. 18 rods of which were electrically heated using tungsten heaters in the rod center. All claddings, corner rods and grid spacers were made of Zr1%Nb (E110) and the shroud of Zr2.5%Nb (E125). For comparison, the QUENCH-06 test (ISP-45) with Western PWR geometry (square lattice) was chosen as reference. QUENCH-12 conducted at the Forschungszentrum Karlsruhe (FZK, Karlsruhe Research Center) on 27 September, 2006 in the frame of the EC-supported ISTC program 1648.2 was proposed by FZK together with RIAR Dimitrovgrad and IBRAE Moscow (Russia), and supported by pretest calculations performed by PSI (Switzerland) and the Kurchatov Institute Moscow (Russia) together with IRSN Cadarache (France). It had been preceded by a low-temperature (maximum 1073 K) pretest on 25 August, 2006 to characterize the bundle thermal hydraulic performance and to provide data to assess the code models used for pretest calculational support. After a stabilization period at 873 K pre-oxidation took place at {proportional_to}1470 K for {proportional_to}3400 s to achieve a maximum oxide thickness of about 200 {mu}m. A transient phase followed with a temperature rise to {proportional_to}2050 K. Then quenching of the bundle by a water flow of 48 g/s was initiated cooling the bundle to ambient temperature in {proportional_to}5 min. Following reflood initiation, a moderate temperature excursion of {proportional_to}50 K was observed, over a longer period than in QUENCH-06. The temperatures at elevations

Results of the QUENCH-12 experiment on reflood of a VVER-type bundle

International Nuclear Information System (INIS)

Stuckert, J.; Grosse, M.; Heck, M.; Schanz, G.; Sepold, L.; Stegmaier, U.; Steinbrueck, M.

2008-09-01

The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor. The QUENCH test bundle with a total length of approximately 2.5 m usually consists of 21 fuel rod simulators of Western PWR (Pressurized Water Reactor) geometry. The QUENCH-12 test bundle, however, which was set up to investigate the effects of VVER materials and bundle geometry (hexagonal lattice) on core reflood consisted of 31 fuel rod simulators. 18 rods of which were electrically heated using tungsten heaters in the rod center. All claddings, corner rods and grid spacers were made of Zr1%Nb (E110) and the shroud of Zr2.5%Nb (E125). For comparison, the QUENCH-06 test (ISP-45) with Western PWR geometry (square lattice) was chosen as reference. QUENCH-12 conducted at the Forschungszentrum Karlsruhe (FZK, Karlsruhe Research Center) on 27 September, 2006 in the frame of the EC-supported ISTC program 1648.2 was proposed by FZK together with RIAR Dimitrovgrad and IBRAE Moscow (Russia), and supported by pretest calculations performed by PSI (Switzerland) and the Kurchatov Institute Moscow (Russia) together with IRSN Cadarache (France). It had been preceded by a low-temperature (maximum 1073 K) pretest on 25 August, 2006 to characterize the bundle thermal hydraulic performance and to provide data to assess the code models used for pretest calculational support. After a stabilization period at 873 K pre-oxidation took place at ∝1470 K for ∝3400 s to achieve a maximum oxide thickness of about 200 μm. A transient phase followed with a temperature rise to ∝2050 K. Then quenching of the bundle by a water flow of 48 g/s was initiated cooling the bundle to ambient temperature in ∝5 min. Following reflood initiation, a moderate temperature excursion of ∝50 K was observed, over a longer period than in QUENCH-06. The temperatures at elevations between 850 mm and 1050 mm exceeded the melting temperature of β-Zr, i

Reflooding Experimental On Beta Test Loop : The Characterisation And Preliminary Experiment

International Nuclear Information System (INIS)

Khairul, H.; Antariksawan, Anhar R.; Sumamo, Edy; Kiswanta; Giarno; Joko, P.; H, Ismu

2001-01-01

The characterisation and preliminary experiment of reflooding had been conducted. The characteristics of main system and component had been identified completely. From these characteristics the experiment condition was set up : heated rod voltage was 20 volt, frequency,of pump was 19 Hz, flow rate was 1 m3/h. The first of experiment did not show the phenomena of rewetting. Possibly because the heated rod temperature was too low. For the second experiment, the voltage of heated rod was increased to 22 Volt and the flow rate was decreased. The result was that the nucleation boiling on the surfaced of heated rod, was observed during the water re flooded the test section

Measuring the socio-economic impacts of agroforestry projects in the Philippines

Science.gov (United States)

Evan Mercer; Belita Vega; Hermie Francisco; Robin Maille

1994-01-01

Conventional wisdom suggests that agroforestry projects can provide both ecological and economic benefits. Most agroforestry project evaluations, however, have failed to adequately assess the soci0-economic impacts. For example, a review of 108 agroforestry project impact evaluations by Sara Scherr of IFPRJ reported that only 8% assessed economic costs or benefits, 5%...

Evaluation and assessment of reflooding models in RELAP5/Mod2.5 and RELAP5/Mod3 codes using Lehigh University and PSI-Neptun bundle experimental data

Energy Technology Data Exchange (ETDEWEB)

Sencar, M.; Aksan, N. [Paul Scherrer Institute, Villigen (Switzerland)

1995-09-01

An extensive analysis and assessment work on reflooding models of RELAP5/Mod2.5 and, RELAP5/Mod3/v5m5 and RELAP/Mod3/v7j have been performed. Experimental data from LehighUniversityv. and PSI-NEPTUN bundle reflooding experiments have been used for the assessment, since both of these tests cover a broad range of initial conditions. Within the range of these initial conditions, it was tried to identify their separate impacts on the calculated results. A total of six Lehigh University reflooding bundle tests and two PSI-NEPTUN tests with bounding initial conditions are selected for the analysis. Detailed nodalisation studies both for hydraulic and conduction heat transfer were done. On the basis of the results obtained from these cases, a base nodalisation scheme was established. All the other analysis work was performed by using this base nodalisation. RELAP5/Mod2.5 results do not change with renodalisation but RELAP5/Mod3 results are more sensitive to renodalisation. The results of RELAP5/Mod2.5 versions show very large deviations from the used experimental data. These results indicate that some of the phenomenology of the events occurring during the reflooding could not be identified. In the paper, detailed discussions on the main reasons of the deviations from the experimental data will be presented. Since, the results and findings of this study are meant to be a developmental aid, some recommendations have been drawn and some of these have already been implemented at PSI with promising results.

Study of top reflooding in case of severe accident and in particular oxidation of Uranium, Zirconium, Oxygen melts; Etude du renoyage par le haut en cas d'accident grave et en particulier oxydation des melanges (U,Zr,O)

Energy Technology Data Exchange (ETDEWEB)

Brunet-Thibault, E

2006-12-15

In 1979, the Three Mile Island (TMI) accident occurred in United States and accelerated research activities in the field of severe accidents. Severe accident management procedures imply massive water injections to flood the core. The work of this thesis bent principally over this reflooding. The first part of the study concerns the core oxidation enhancement during the reflooding phase which leads to a rough increase of the concentration of burnable hydrogen in the containment. This is why the study carried on the analysis of the contribution of the oxidation of U-Zr-O mixtures, towards the total production of hydrogen during reflooding. In the second part, the study concerns top flooding modelling i.e.: with injection of water in the hot legs. Here, we attempted to define bases and realize a model allowing to describe this type of reflooding. These models were validated on the simulation of the parameter with MAAP4 code. (author)

Lepidoptera, Nymphalidae, Heliconiinae, Heliconius sara apseudes (Hübner, 1813: Distribution extension

Directory of Open Access Journals (Sweden)

Iserhard, C. A.

2010-01-01

Full Text Available This work presents new records and extends the geographic distribution of Heliconius sara apseudes in theAtlantic Forest of the state of Rio Grande do Sul. Five new records were taken along butterfly inventories carried outbetween 2005 and 2010 in distinct phytophysiognomies at Rio Grande do Sul northeast region: Swamp Forest, AtlanticForest stricto sensu and Araucaria Moist Forest. The fact that all registers occurred in well preserved habitats of the AtlanticForest emphasizes the need of conservation of this biome in Rio Grande do Sul.

Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

International Nuclear Information System (INIS)

Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

2017-01-01

A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

Energy Technology Data Exchange (ETDEWEB)

Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

International Nuclear Information System (INIS)

Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

1983-09-01

TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

Evaluation report on CCTF core-I reflood test C1-5 (Run 14)

International Nuclear Information System (INIS)

Murao, Yoshio; Akimoto, Hajime; Sudoh, Takashi; Okubo, Tsutomu

1983-02-01

A study of a cylindrical core test facility (CCTF) test was performed for modeling the system behavior during the reflood phase of a PWR-LOCA and the following conclusions were obtained: 1) With the exception of some points, the observed phenomena are similar to a model derived from an evaluation model for a PWR safety evaluation. 2) The different points are the water accumulation in the upper plenum, the ECC bypass in the downcomer, the reduction of the effective downcomer head and the pressure drop at the broken cold leg nozzle and in the interconnected pipes. (author)

Study of top reflooding in case of severe accident and in particular oxidation of Uranium, Zirconium, Oxygen melts; Etude du renoyage par le haut en cas d'accident grave et en particulier oxydation des melanges (U,Zr,O)

Energy Technology Data Exchange (ETDEWEB)

Brunet-Thibault, E

2006-12-15

In 1979, the Three Mile Island (TMI) accident occurred in United States and accelerated research activities in the field of severe accidents. Severe accident management procedures imply massive water injections to flood the core. The work of this thesis bent principally over this reflooding. The first part of the study concerns the core oxidation enhancement during the reflooding phase which leads to a rough increase of the concentration of burnable hydrogen in the containment. This is why the study carried on the analysis of the contribution of the oxidation of U-Zr-O mixtures, towards the total production of hydrogen during reflooding. In the second part, the study concerns top flooding modelling i.e.: with injection of water in the hot legs. Here, we attempted to define bases and realize a model allowing to describe this type of reflooding. These models were validated on the simulation of the parameter with MAAP4 code. (author)

Reflooding phase of the LOCA - state of the art II. Rewetting and liquid entrainment

International Nuclear Information System (INIS)

Elias, E.; Yadigaroglu, G.

1977-01-01

Understanding the mechanisms by which hot fuel rods quench and the physics of liquid droplet entrainment is important for the analysis of the reflooding phase of the LOCA. Published models of the rewetting process include simple one-dimensional solutions. The basic physical assumptions of these models and the numerical values assigned to the various parameters, as well as empirical rewetting correlations are discussed. The various mechanisms for liquid droplet entrainment and analytical formulations of the critical gas velocity and of the droplet diameter at the onset of entrainment are reviewed

Measurement of two-phase flow at the core upper plenum interface under simulated reflood conditions

International Nuclear Information System (INIS)

Thomas, D.G.; Combs, S.K.; Bagwell, M.E.

1980-01-01

Objectives of the Instrument Development Loop program were to simulate flows at the core/upper plenum interface during the reflood phase of a LOCA and to develop instruments for measuring mass-flows at this interface. A tie plate drag body was developed and tested successfully, and the data obtained were shown to be equivalent to pressure drops. The tie-plate drag body gave useful measurements in pure downflow, and the drag/turbine combination correlates with mass flow for high upflow

The blowdown, refill and reflood phase during a LOCA. Survey of the main physical phenomena

International Nuclear Information System (INIS)

Reocreux, M.

1980-05-01

In this paper, the main physical phenomena occuring during a LOCA are reviewed. They are presented in a chronological order. For each phenomena, a detailed physical description is given followed by the review of the general modelling problems. For some of these phenomena, modelling details are given for critical flow, for two-phase flow and heat transfer, for critical heat flux and post critical heat flux heat transfer, for reflood and rewet heat transfer and in the survey on LOCA computation codes

REFLA-1D/MODE 1: a computer program for reflood thermo-hydrodynamic analysis during PWR-LOCA user's manual

International Nuclear Information System (INIS)

Murao, Yoshio; Sugimoto, Jun; Okubo, Tsutomu

1981-01-01

This manual describes the REFLA-1D/MODE 1 reflood system analysis code. This code can solve the core thermo-hydrodynamics under forced flooding conditions and gravity feed conditions in a system similar to FLECHT-SET phase A. This manual describes the REFLA-1D/MODE 1 models and provides application information required to utilize REFLA-1D/MODE 1. (author)

The staphylococcal accessory regulator, SarA, is an RNA-binding protein that modulates the mRNA turnover properties of late-exponential and stationary phase Staphylococcus aureus cells

Directory of Open Access Journals (Sweden)

John M Morrison

2012-03-01

Full Text Available The modulation of mRNA turnover is gaining recognition as a mechanism by which Staphylococcus aureus regulates gene expression, but the factors that orchestrate alterations in transcript degradation are poorly understood. In that regard, we previously found that 138 mRNA species, including the virulence factors protein A (spa and collagen binding protein (cna, are stabilized in a sarA-dependent manner during exponential phase growth, suggesting that SarA protein may directly or indirectly effect the RNA turnover properties of these transcripts. Herein, we expanded our characterization of the effects of sarA on mRNA turnover during late exponential and stationary phases of growth. Results revealed that the locus affects the RNA degradation properties of cells during both growth phases. Further, using gel mobility shift assays and RIP-ChIP, it was found that SarA protein is capable of binding mRNA species that it stabilizes both in vitro and within bacterial cells. Taken together, these results suggest that SarA post-transcriptionally regulates S. aureus gene expression in a manner that involves binding to and consequently altering the mRNA turnover properties of target transcripts.

Computation of 3D thermohydraulics in partially blocked bundles during the reflood phase of a LOCA

International Nuclear Information System (INIS)

Cicero, G.M.; Briere, E.; Fornaciari, G.

1994-06-01

In Pressurized Water Reactors (PWR), ballooning of the fuel rod claddings may occur during a LOCA, since the fuel rod claddings are heated up, and the system pressure is low. The severe blockages that may result induce cross-flow diversion and three-dimensional effects on thermohydraulics in the core bundle, during the reflood phase. To improve the knowledge of these phenomena and their physical modelling in the code CATHARE, 3D computer codes are needed. In 1990, EDF has started up a development and validation program of the 3D THYC computer code to analyze the thermohydraulics of the flow during the reflood phase, in partially blocked bundles. The main objective is to calculate the temperatures of the rods above the quench front, when they are cooled by superheated steam with saturated droplets. First, this paper introduces the THYC model developed for reflood studies. Secondly, we report the first qualification results on a Flooding Experiments with Blocked Array (FEBA) test. Thirdly, we analyze the model predictions on a large break LOCA transient, in a 900 MW PWR 11x11 core area with a 3x3 central blockage. THYC simulates the transient in the bundle around and above the blockage, until the quench front enters the computational domain. Previously, a 1D CATHARE simulation gives the boundary conditions and, in the reactor core case, the deformation of the blocked fuel rods. The results analysis focused on the time evolution of the clad temperatures in the blocked and in the bypass region. In the FEBA test simulation, the main observations are properly predicted within the blockage. Temperatures are lower in blocked rod sleeves than in unblocked rod claddings since the steam gap reduces the power transmitted by the heater rod to the sleeve. In the core case, the model predicts the opposite result. Within the blockage, ballooned rod temperatures are higher than non-ballooned rod ones. We show by sensitivity studies that these behaviour difference between FEBA rods

Intravenous immunoglobulin in the management of a rare cause of hemolytic disease of the newborn: Anti-SARA antibodies.

Science.gov (United States)

Venkataraman, Rohini; Yusuf, Kamran

2017-01-01

Hemolytic disease of newborn (HDN) is a condition that develops in a fetus, when the IgG molecules produced by the mother pass through the placenta and attack the fetal red blood cells. HDN can occur due to Rh and ABO incompatibilities between the mother and the fetus as well as due to other allo-immune antibodies belonging to Kell (K and k), Duffy (Fya), Kidd (Jka and Jkb), and MNS (M, N, S, and s) systems. Role of intravenous immunoglobulin in management of HDN is not clear.SARA red blood cell antigen, first discovered in 1990 is a low frequency antigen. We report, a multiparous female whose pregnancy was complicated by HDN due to anti-SARA antibodies requiring both exchange transfusion and intravenous immunoglobulin. The response was sustained after intravenous immunoglobulin (IVIG) rather than after exchange transfusion.

LDA measurement of droplet behavior across tie plate during dispersed flow portion of loca reflood

International Nuclear Information System (INIS)

Lee, S.L.; Srinivasan, J.; Cho, S.K.

1980-01-01

The flow of an air-water droplet dispersion in a simulated 3-D test section in the reflood portion of LOCA was studied. For this purpose, a new scheme of Laser-Doppler Anemometry for the simultaneous measurement of size and velocity of large-size [0.5 mm-6 mm] droplets was developed and utilized. It was observed that the size distribution of the reentrained droplets depends mainly on the flow regimes and is essentially independent of that of the incoming dispersion below the tie plate. 8 refs

Evaluation report on CCTF CORE-I REFLOOD TEST Cl-4 (Run 13) and Cl-15 (Run 24)

International Nuclear Information System (INIS)

Sudoh, Takashi; Murao, Yoshio.

1983-08-01

The tests Cl-4 and Cl-15 were performed with the Cylindrical Core Test Facility (CCTF) to investigate the effects of the depressurization process to simulate the refill phase, and the effects of the nitrogen to be injected after the end of the accumulator injection on the thermo-hydraulic behavior in the core and primary loop system during refill and reflood phases. In these tests, after the lower plenum was filled to 0.9m level with saturated water at 0.6 MPa, the accumulator water was injected into three intact cold legs in the depressurization period from 0.6 MPa to 0.2 MPa. The water in the lower plenum voided during the depressurization and the significant steam condensation occurred in or near the intact cold legs. The condensation caused high steam flow rate in the intact loops and the lower plenum flashing resulted in suppressed core water accumulation. The slightly lower core heat transfer coefficient due to the less core water caused the higher turnaround temperature and the longer quench time than those of the normal reflood test without the depressurization process. The nitrogen injection followed the accumulator injection was allowed in the test Cl-15. However, significant effects of the nitrogen injection was not observed. (author)

Evaluation report on CCTF core-I reflood test C1-19 (RUN 38)

International Nuclear Information System (INIS)

Murao, Yoshio; Fujiki, Kazuo; Akimoto, Hajime

1983-02-01

A test named the Evaluation Model (EM) test was performed, whose test conditions were simulated the reflood phase predicted with the safety evaluation analysis. The test results were compared with the blindfold results predicted by Evaluation Model (EM) codes. The main conclusions are as follows: (1) The core heat transfer model built in the EM codes gives conservative results. (2) The system models in the present EM codes are found to be well balanced integrally over the system. (3) Conservative items and items to be improved are pointed out. The downcomer slow water accumulation observed in the lower flow rate test was not appeared in the EM test. (author)

Statistical Uncertainty Quantification of Physical Models during Reflood of LBLOCA

Energy Technology Data Exchange (ETDEWEB)

Oh, Deog Yeon; Seul, Kwang Won; Woo, Sweng Woong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

The use of the best-estimate (BE) computer codes in safety analysis for loss-of-coolant accident (LOCA) is the major trend in many countries to reduce the significant conservatism. A key feature of this BE evaluation requires the licensee to quantify the uncertainty of the calculations. So, it is very important how to determine the uncertainty distribution before conducting the uncertainty evaluation. Uncertainty includes those of physical model and correlation, plant operational parameters, and so forth. The quantification process is often performed mainly by subjective expert judgment or obtained from reference documents of computer code. In this respect, more mathematical methods are needed to reasonably determine the uncertainty ranges. The first uncertainty quantification are performed with the various increments for two influential uncertainty parameters to get the calculated responses and their derivatives. The different data set with two influential uncertainty parameters for FEBA tests, are chosen applying more strict criteria for selecting responses and their derivatives, which may be considered as the user’s effect in the CIRCÉ applications. Finally, three influential uncertainty parameters are considered to study the effect on the number of uncertainty parameters due to the limitation of CIRCÉ method. With the determined uncertainty ranges, uncertainty evaluations for FEBA tests are performed to check whether the experimental responses such as the cladding temperature or pressure drop are inside the limits of calculated uncertainty bounds. A confirmation step will be performed to evaluate the quality of the information in the case of the different reflooding PERICLES experiments. The uncertainty ranges of physical model in MARS-KS thermal-hydraulic code during the reflooding were quantified by CIRCÉ method using FEBA experiment tests, instead of expert judgment. Also, through the uncertainty evaluation for FEBA and PERICLES tests, it was confirmed

Experimental results of the effective water head in downcomer during reflood phase of a PWR LOCA

International Nuclear Information System (INIS)

Sudo, Yukio; Murao, Yoshio; Akimoto, Hajime

1980-08-01

The results and analysis of an experiment for the effective water head in downcomer with 50mm gap size are described. The main objective of the experiment was to clarify the effect of gap size on reflooding in a PWR LOCA. The effective water head in downcomer is the driving force for feeding emergency coolant into the core during reflood phase of a PWR LOCA. Discussions presented here follow those of a previous report in which experimental results and analysis were described for the case of 200mm gap size. Experimental Conditions were: Initial Wall Temperature = 200 -- 300 0 C, Back Pressure = 1 atm., Coolant Temperature = 71 -- 100 0 C, Extraction Water Velocity = 0 -- 2 cm/s, Gap Size = 50 mm. The effective water head history obtained in the experiment was compared with those predicted with Sudo's void fraction correlation. In the prediction, heat input to coolant was calculated from the response of measured wall temperature with heat condition analysis. The experimental results and analysis reveals that: (1) The effects of the gap size and initial wall temperature are evident, (2) The effect of extraction water velocity is negligible, and (3) The predicted history of effective water head is in good agreement with the experimental results except during the transient period in which the effective water head is descreasing. (author)

Investigation of typicality of non-nuclear rod and fuel-clad gap effect during reflood phase, and development of a FEM thermal transient analysis code HETFEM

International Nuclear Information System (INIS)

Sudoh, Takashi

1981-06-01

The objective of this study are: 1) Evaluate the capability of the electrical heater for simulating the fuel rod during the reflood phase, and 2) To investigate the effect of the clad-fuel gap in the fuel rod on the clad thermal response during the reflood phase. A computer code HETFEM which is the two dimensional transient thermal conductivity analysis code utilized a finite element method is developed for analysing thermal responses of heater and fuel rod. The two kinds of electrical heaters and a fuel rod are calculated with simple boundary conditions. 1) direct heater (former JAERI reflood test heater), 2) indirect heater (FLECHT test heater), 3) fuel rod (15 x 15 type in Westinghouse PWR). The comparison of the clad temperature responses shows the quench time is influenced by the thermal diffusivity and gap conductance. In the conclusion, the ELECHT heater shows atypicality in the clad temperature response and heat releasing rate. But the direct heater responses are similar to those of the fuel rod. For the gap effect on the fuel rod behavior, the lower gap conductance causes sooner quench and less heat releasing rate. This calculation is not considered the precursory cooling which is affected by heat releasing rate at near and below the quench front. Therefore two dimensional calculation with heat transfer related to the local fluid conditions will be needed. (author)

Reading, Laterality, and the Brain: Early Contributions on Reading Disabilities by Sara S. Sparrow

OpenAIRE

Fletcher, Jack M.; Morris, Robin D.

2014-01-01

Although best known for work with children and adults with intellectual disabilities and autism spectrum disorders, training in speech pathology and a doctorate in clinical psychology and neuropsychology was the foundation for Sara Sparrow’s long-term interest in reading disabilities. Her first papers were on dyslexia and laterality, and the maturational lag theory of developmental dyslexia proposed with Paul Satz, her mentor. The research program that emerged from this work had a wide impact...

Investigation of radial power and temperature effects in large-scale reflood experiments

International Nuclear Information System (INIS)

Motley, F.

1983-01-01

The largest reflood test facility in the world has been designed and constructed by the Japan Atomic Energy Research Institute (JAERI). The experimental test facility, known as the Cylindrical Core Test Facility (CCTF), models a full-height core section and the four primary loops of a Pressurized Water Reactor (PWR). The radial power distribution and temperature distribution were varied during the testing program. The test results indicate that the radial effects, while noticeable, do not appreciably alter the overall quenching behavior of the facility. The Transient Reactor Analysis Code (TRAC) correctly predicted the experimental results of several of the tests. The code results indicate that the core flow pattern adjusts multidimensionally to mitigate the effects of increased power or stored energy

Core-debris quenching-heat-transfer rates under top- and bottom-reflood conditions

International Nuclear Information System (INIS)

Ginsberg, T.; Tutu, N.; Klages, J.; Schwarz, C.E.; Sanborn, Y.

1983-02-01

This paper presents recent experimental data for the quench-heat-transfer characteristics of superheated packed beds of spheres which were cooled, in separate experiments, by top- and bottom-flooding modes. Experiments were carried out with beds of 3-mm steel spheres of 330-mm height. The initial bed temperature was 810 K. The observed heat-transfer rates are strongly dependent on the mode of water injection. The results suggest that top-flood bed quench heat transfer is limited by the rate at which water can penetrate the bed under two-phase countercurrent-flow conditions. With bottom-reflood the heat-transfer rate is an order-of-magnitude greater than under top-flood conditions and appears to be limited by particle-to-fluid film boiling heat transfer

Should Community College Be Free? Forum. "Education Next" Talks with Sara Goldrick-Rab and Andrew P. Kelly

Science.gov (United States)

Goldrick-Rab, Sara; Kelly, Andrew P.

2016-01-01

In this article, "Education Next" talks with Sara Goldrick-Rab and Andrew Kelly. President Obama's proposal for tuition-free community college, seems to have laid down a marker for the Democratic Party. Vermont senator Bernie Sanders is touting his plan for free four-year public college on the primary trail; Massachusetts senator…

Assessment of TRAC-BF1 1D reflood model with CCTF and SCTF data

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

1993-03-01

Post test calculations for six selected Cylindrical Core Test Facility (CCTF) and Slab Core Test Facility (SCTF) tests were performed to assess the core thermal hydraulic models of the TRAC-BF1 code during the reflood in a PWR LOCA. A special version of the TRAC code was developed at JAERI by implementing the constitutive package of the TRAC-BF1 code into the TRAC-PF1 code for this assessment. The TRAC-BF1 model predicted well the void fraction at either bottom or top part of the core and overpredicted the void fraction at the center part of the core in the CCTF and SCTF tests performed under so-called licensing conditions. The TRAC-BF1 model overpredicted the clad temperatures at the center part of the core. The TRAC-BF1 model predicted a jump of void fraction where the flow pattern transition between the bubbly/slug flow and the annular/dispersed flow regimes occurred. The jump caused the water mass flow rate to be unstable and resulted in the overprediction of the void fraction at the center part of the core. It was also found that the TRAC-BF1 film boiling model underestimated the heat transfer coefficient in the vicinity of the quench front and caused the quench front propagation to be delayed. These assessment results suggest the following areas should be improved in future to apply the TRAC-BF1 code to the reflood in a PWR LOCA: (1) Core hydraulic model where flow pattern transition occurs, (2) Core heat transfer model in the film boiling regime, especially for the dependence on the distance from the quench front. (author)

Effects of upper plenum injection on thermo-hydrodynamic behavior under refill and reflood phases

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Abe, Yutaka; Adachi, Hiromichi; Ohnuki, Akira; Osakabe, Masahiro

1984-12-01

In order to investigate the thermo-hydrodynamic behavior in core under simultaneous ECC water injection into the upper plenum and the intact cold leg during the refill and reflood phases of a PWR-LOCA, Tests S1-SH3 and S1-SH4 were performed by using Slab Core Test Facility (SCTF) with the injection of saturated and 67K subcooled water into the upper plenum, respectively, under the same cold leg injection condition. The following major findings were obtained by examining these test results. (1) Although the core was cooled by the fall back water from the upper plenum into the core during the period of high injection rate into the upper plenum, the core was cooled mainly by the bottom flooding after the BOCREC (Bottom of core recovery). (2) The possible fall back flow rate estimated with a CCFL correlation rapidly decreased after the BOCREC because of the increase of steam generation rate in core. (3) Continuous fall back of subcooled water was not observed even under the condition with large upper plenum injection rate of subcooled water and with steam outflow through the lower plenum into the downcomer. The fall back was intermittently limited by the rapid increase of upward steam flow which was generated in the core due to the evaporation of the fall back water. (4) The rising of liquid level in the lower plenum was suppressed by the pressurization in core due to the evaporation of fall back water before the BOCREC and therefore the beginning of bottom reflood was delayed. Some selected data from Tests S1-SH3 and S1-SH4 are also included in this report. (author)

Analysis with SCDAP/RELAP5 of reflooding of an overheated core in Forsmark 3 BWR after loss of electric power

International Nuclear Information System (INIS)

Nilsson, Lars.

1993-01-01

In foregoing SIK-2.2 project two severe accident sequences for Forsmark 3 comprising loss of electric power (Total Blackout, TB) without recovery actions (e.g. emergency cooling) were analysed with SCDAP/RELAP5. Code version MOD2.5/V3f was applied and a single core channel input model was used. Present analysis was done with the same initiating events as in SIK-2.2, but assuming recovery of auxiliary feed water and/or emergency core cooling systems to take place after heat-up of the core, in compliance with the plans of the SIK-2.3 project. A new test version of the SCDAP/RELAP5 code, version MOD3/V7af, was used here. The geometrical input model was extended from having one, into five parallel core zones, still with ten axial core sub volumes. Calculations were performed for three TB cases, one without cooling recovery, and two with injection of cold water at a rate of 45 kg/s, beginning when the maximum cladding temperature has reached 1522 K. The results show that a considerably more efficient cooling was obtained spraying the water to the top of the core, compared to the case where the cooling water was introduced into the downcomer, leading to bottom reflooding

Full-scale model development of the WWER-440 reactor fuel rod bundle for core temperature regime study under reflooding conditions

International Nuclear Information System (INIS)

Bezrukov, Yu.A.; Logvinov, S.A.; Levchuk, S.V.; Nakladnov, V.D.; Onshin, V.P.; Sokolov, A.S.

1982-01-01

Consideration is given to the issues of a full scale WWER-440 fuel rod bundle imitation. An imitator contains a molybdenum heating rod inclosed in stainless steel shell. The shell diameter is 9 mm, the heated length is 2500 mm, the total len.o.th is 2855 mm. 125 fuel rod imitators are set in the bundle mock-up. The experiments were run on a test facility imitating the WWER-440 reactor primary loop, providing the conditions of the loop breaking. The mock-up thermal hydraulics has been studied during the refloodino. stage. The mock-up was heated up to predetermined initial temperature at a low power level with saturated steam cooling. Then the steam input was stopped, the power level rarapidly rised up to a given value and the cooling water injected. Simultaneously with water injection all the measured parameters monitoring was started. Both at the top spraying and combined cooling temperature oscillations in the upper and middle parts of the mock-up were observed. At the bottom reflooding the mock-up cooling down took more time, thereat temperature inthe upper part first slowly rised during reflooding then decreased and then dropped abruptly at thefront coming up [ru

Assessment of MARS for downcomer multi-dimensional thermal hydraulics during LBLOCA reflood using KAERI air-water direct vessel injection tests

Energy Technology Data Exchange (ETDEWEB)

Won-Jae, Lee; Kwi-Seok, Ha; Chul-Hwa, Song [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

2001-07-01

The MARS code has been assessed for the downcomer multi-dimensional thermal hydraulics during a large break loss-of-coolant accident (LBLOCA) reflood of Korean Next Generation Reactor (KNGR) that adopted an upper direct vessel injection (DVI) design. Direct DVI bypass and downcomer level sweep-out tests carried out at 1/50-scale air-water DVI test facility are simulated to examine the capability of MARS. Test conditions are selected such that they represent typical reflood conditions of KNGR, that is, DVI injection velocities of 1.0 {approx} 1.6 m/sec and air injection velocities of 18.0 {approx} 35.0 m/sec, for single and double DVI configurations. MARS calculation is first adjusted to the experimental DVI film distribution that largely affects air-water interaction in a scaled-down downcomer, then, the code is assessed for the selected test matrix. With some improvements of MARS thermal-hydraulic (T/H) models, it has been demonstrated that the MARS code is capable of simulating the direct DVI bypass and downcomer level sweep-out as well as the multi-dimensional thermal hydraulics in downcomer, where condensation effect is excluded. (authors)

Preliminary analysis of the effect of the grid spacers on the reflood heat transfer

International Nuclear Information System (INIS)

Sugimoto, Jun; Murao, Yoshio

1982-02-01

The results are described about the preliminary analysis of the effect of the grid spacers on the heat transfer during reflood phase of a PWR LOCA. Experiments at JAERI and other facilities showed substantial heat transfer enhancement near the grid spacers. The heat transfer enhancement decreases with the distance from the grid spacers in the downstream region of the grid spacers. Several mechanisms are discussed about the heat transfer enhancement near the grid spacers. A model of a coalescence of the water droplets downstream the spacers is proposed based on the review of the experimental data. The heat transfer correlation for the saturated film boiling is utilized to quantify the heat transfer augmentation by the grid spacers. (author)

Multi-dimensional Mixing Behavior of Steam-Water Flow in a Downcomer Annulus during LBLOCA Reflood Phase with a DVI Injection Mode

International Nuclear Information System (INIS)

Kwon, T.S.; Yun, B.J.; Euh, D.J.; Chu, I.C.; Song, C.H.

2002-01-01

Multi-dimensional thermal-hydraulic behavior in the downcomer annulus of a pressurized water reactor vessel with a Direct Vessel Injection (DVI) mode is presented based on the experimental observation in the MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) steam-water test facility. From the steady-state test results to simulate the late reflood phase of a Large Break Loss-of-Coolant Accidents(LBLOCA), isothermal lines show the multidimensional phenomena of a phasic interaction between steam and water in the downcomer annulus very well. MIDAS is a steam-water separate effect test facility, which is 1/4.93 linearly scaled-down of 1400 MWe PWR type of a nuclear reactor, focused on understanding multi-dimensional thermalhydraulic phenomena in downcomer annulus with various types of safety injection during the refill or reflood phase of a LBLOCA. The initial and the boundary conditions are scaled from the pre-test analysis based on the preliminary calculation using the TRAC code. The superheated steam with a superheating degree of 80 K at a given downcomer pressure of 180 kPa is injected equally through three intact cold legs into the downcomer. (authors)

Sara, a patient with borderline personality structure: A "crisis" management

Directory of Open Access Journals (Sweden)

Paola Marinelli

2015-05-01

Full Text Available In this article the author examines some significant passages of his treatment of a patient with borderline personality structure, with the intention of giving a formative contribution to the delicate issue of the search for congruence between theory and clinic operations. This reflection is therefore an opportunity to integrate these aspects. The individualization of the therapeutic relationship in the theoretical framework of group analysis allowed the emotional investment in the person of the therapist, which is useful in the construction of a meaningful relationship on the human, emotional and cognitive plane; a space within which it has become increasingly possible for Sara, share and process emotions, re-build, contact parts of the self frustrated and disappointed, perceive less and less the void and become less vulnerable, being able to pull over to the original trauma. 

A Different Curriculum of Preparation for Work: Commentary on Mike Rose, Sara Goldrick-Rab, Kris Gutierrez and Norton Grubb

Science.gov (United States)

Worthen, Helena Harlow

2012-01-01

The January 2012 issue of "Mind, Culture, and Activity" published the Invited Presidential Address "Rethinking Remedial Education and the Academic-Vocational Divide," given by Mike Rose at the 2011 meeting of the American Educational Research Association in New Orleans, along with responses and commentary by Sara Goldrick-Rab, Kris Gutierrez, and…

Evaluation report on CCTF Core-II reflood Test C2-15 (Run 75)

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Akimoto, Hajime; Murao, Yoshio

1992-01-01

This report presents an evaluation on the CCTF Core-II Test C2-15 (Run 75). The purpose of the test is to investigate whether the thermo-hydrodynamic behavior is different between the CCTF and the FLECHT-SET reflooding tests. For this purpose, test conditions of the present test were set as close as possible to those of concerned FLECHT-SET 2714B experiment, taking account of differences in facility design. Investigating results of both the tests, the following conclusions are obtained: (1) Some discrepancies were observed in the measured test conditions between the two tests. Out of them, difference in the Acc injection duration was large and affected test results, such as the water accumulation in the downcomer and the core and the core cooling, during the initial period. However, this effect was found to become small with time. (2) Taking account of this difference and the difference in the broken cold leg pressure loss coefficient between the two facilities, the overall reflooding behavior is judged to be similar in the two facilities. (3) The CCTF test results showed the core heat transfer enhancement in the higher power region due to its steep radial power distribution, whereas the FLECHT-SET did not due to its rather flat radial power distribution. This enhancement was observed significantly at 1.83 m but was smaller at the higher elevation. (4) The heat transfer was nearly identical between the two tests and an existing correlation could well predict the heat transfer coefficients of both the tests at the location where the heat transfer enhancement mentioned above (3) were small, during the time period when the effect of the difference in the Acc injection mentioned above (1) were small. (5) Therefore, the core cooling is expected to be almost the same in the CCTF and the FLECHT-SET under the same core boundary conditions and core radial power distribution. (author)

Analytical model for bottom reflooding heat transfer in light water reactors (the UCFLOOD code)

International Nuclear Information System (INIS)

Arrieta, L.; Yadigaroglu, G.

1978-08-01

The UCFLOOD code is based on mechanistic models developed to analyze bottom reflooding of a single flow channel and its associated fuel rod, or a tubular test section with internal flow. From the hydrodynamic point of view the flow channel is divided into a single-phase liquid region, a continuous-liquid two-phase region, and a dispersed-liquid region. The void fraction is obtained from drift flux models. For heat transfer calculations, the channel is divided into regions of single-phase-liquid heat transfer, nucleate boiling and forced-convection vaporization, inverted-annular film boiling, and dispersed-flow film boiling. The heat transfer coefficients are functions of the local flow conditions. Good agreement of calculated and experimental results has been obtained. A code user's manual is appended

Impact of the Regulators SigB, Rot, SarA and sarS on the Toxic Shock Tst Promoter and TSST-1 Expression in Staphylococcus aureus.

Directory of Open Access Journals (Sweden)

Diego O Andrey

Full Text Available Staphylococcus aureus is an important pathogen manifesting virulence through diverse disease forms, ranging from acute skin infections to life-threatening bacteremia or systemic toxic shock syndromes. In the latter case, the prototypical superantigen is TSST-1 (Toxic Shock Syndrome Toxin 1, encoded by tst(H, and carried on a mobile genetic element that is not present in all S. aureus strains. Transcriptional regulation of tst is only partially understood. In this study, we dissected the role of sarA, sarS (sarH1, RNAIII, rot, and the alternative stress sigma factor sigB (σB. By examining tst promoter regulation predominantly in the context of its native sequence within the SaPI1 pathogenicity island of strain RN4282, we discovered that σB emerged as a particularly important tst regulator. We did not detect a consensus σB site within the tst promoter, and thus the effect of σB is likely indirect. We found that σB strongly repressed the expression of the toxin via at least two distinct regulatory pathways dependent upon sarA and agr. Furthermore rot, a member of SarA family, was shown to repress tst expression when overexpressed, although its deletion had no consistent measurable effect. We could not find any detectable effect of sarS, either by deletion or overexpression, suggesting that this regulator plays a minimal role in TSST-1 expression except when combined with disruption of sarA. Collectively, our results extend our understanding of complex multifactorial regulation of tst, revealing several layers of negative regulation. In addition to environmental stimuli thought to impact TSST-1 production, these findings support a model whereby sporadic mutation in a few key negative regulators can profoundly affect and enhance TSST-1 expression.

Construct Validity and Reliability of the SARA Gait and Posture Sub-scale in Early Onset Ataxia

Directory of Open Access Journals (Sweden)

Tjitske F. Lawerman

2017-12-01

Full Text Available Aim: In children, gait and posture assessment provides a crucial marker for the early characterization, surveillance and treatment evaluation of early onset ataxia (EOA. For reliable data entry of studies targeting at gait and posture improvement, uniform quantitative biomarkers are necessary. Until now, the pediatric test construct of gait and posture scores of the Scale for Assessment and Rating of Ataxia sub-scale (SARA is still unclear. In the present study, we aimed to validate the construct validity and reliability of the pediatric (SARAGAIT/POSTURE sub-scale.Methods: We included 28 EOA patients [15.5 (6–34 years; median (range]. For inter-observer reliability, we determined the ICC on EOA SARAGAIT/POSTURE sub-scores by three independent pediatric neurologists. For convergent validity, we associated SARAGAIT/POSTURE sub-scores with: (1 Ataxic gait Severity Measurement by Klockgether (ASMK; dynamic balance, (2 Pediatric Balance Scale (PBS; static balance, (3 Gross Motor Function Classification Scale -extended and revised version (GMFCS-E&R, (4 SARA-kinetic scores (SARAKINETIC; kinetic function of the upper and lower limbs, (5 Archimedes Spiral (AS; kinetic function of the upper limbs, and (6 total SARA scores (SARATOTAL; i.e., summed SARAGAIT/POSTURE, SARAKINETIC, and SARASPEECH sub-scores. For discriminant validity, we investigated whether EOA co-morbidity factors (myopathy and myoclonus could influence SARAGAIT/POSTURE sub-scores.Results: The inter-observer agreement (ICC on EOA SARAGAIT/POSTURE sub-scores was high (0.97. SARAGAIT/POSTURE was strongly correlated with the other ataxia and functional scales [ASMK (rs = -0.819; p < 0.001; PBS (rs = -0.943; p < 0.001; GMFCS-E&R (rs = -0.862; p < 0.001; SARAKINETIC (rs = 0.726; p < 0.001; AS (rs = 0.609; p = 0.002; and SARATOTAL (rs = 0.935; p < 0.001]. Comorbid myopathy influenced SARAGAIT/POSTURE scores by concurrent muscle weakness, whereas comorbid myoclonus predominantly influenced

Development of an ion guide coupled to an on-line isotope separation system on Sara. Identification and study of isospin exotic nuclei at Isolde and Sara

International Nuclear Information System (INIS)

Bouldjedri, A.

1992-06-01

This work is concerned with the study of exotic nuclei located on both sides of the stability-line and known as neutron rich and neutron deficient respectively. For the former, produced by alpha particle-induced fission, an on-line isotope separation with an ion guide (IGISOL) has been developed and submitted to several off-line and on-line optimization tests showing capacity to spectroscopic studies. In the case of neutron deficient nuclei near the magicity Z=82, 182 Tl(3s) has been identified and its decaying modes and those of 183 Tl ground state, studied, using the on-line separator ISOLDE. On the other hand, the β decay of 172,175 Ir produced in 32 S induced reaction is studied using a helium jet system on the SARA accelerator. Existence of isomers is derived from half-lives measurements

Evaluation report on CCTF Core-I reflood tests C1-5 (Run 14), C1-7 (Run 16) and C1-14 (Run 23)

International Nuclear Information System (INIS)

Sugimoto, Jun; Muurao, Yoshio

1983-02-01

The present report describes the effects of the initial clad temperature on the reflood phenomena observed in the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute. The evaluation is based on the data of tests C1-5, C1-7 and C1-14 of the CCTF-Core I test series. Nominal initial peak clad temperatures in these tests are 600 0 C, 700 0 C and 800 0 C, respectively. With the higher initial clad temperature, the higher loop mass flow rate and the lower water accumulation in the core and the upper plenum were obtained in an early reflood transient. However, the core inlet flow conditions, which is sensitive to the core cooling, were not much affected by the higher initial clad temperature. The slower quench front propagation was observed with the higher initial clad temperature. However, the heat transfer coefficient was almost identical with each other before the turnaround time, which resulted in the lower temperature rise with the highest initial clad temperature. This qualitatively agreed with the results of the forced feed FLECHT experiment. (author)

Assessment of TRAC-PD2 reflood core thermo-hydraulic model by CCTF Test C1-16

International Nuclear Information System (INIS)

Sugimoto, Jun

1982-11-01

The TRAC-PD2 reflood core thermo-hydraulic model was assessed by CCTF Test C1-16. The measured data were utilized as core boundary conditions in the TRAC calculations. The results indicate that the core inlet liquid temperature and the core heater rod temperatures are in reasonable agreement with data, but the pressure distribution in the core and water pool formation in the upper plenum are not in good agreement. The parametric effects of the droplet critical Weber number, the material properties of the heater rod, the noding of the upper plenum, and the minimum stable film boiling temperature are also discussed. (author)

Reflooding phase of the LOCA in PWRs. Part II: rewetting and liquid entrainment

International Nuclear Information System (INIS)

Elias, E.; Yadigaroglu, G.

1978-01-01

Surface rewetting and liquid-droplet entrainment play an important role in the analysis of the reflooding phase of the loss-of-coolant accident in pressurized-water reactors. The definitions and the various interpretations given to the rewetting temperature and the rewetting mechanisms of the fuel rods are discussed. Published models of the axial-conduction-controlled rewetting process include one-dimensional solutions in two axial regions, one-dimensional solutions in three axial regions with or without precursory cooling, one- and two-dimensional numerical-difference techniques using temperature-dependent heat-transfer coefficients, and analytical two-dimensional solutions. The basic physical assumptions and the numerical values assigned to the various parameters, as well as empirical rewetting correlations, are discussed. The physical mechanisms for liquid-droplet entrainment and analytical formulations of the critical gas velocity and of the droplet diameter at the onset of entrainment are reviewed

Experiment predictions of LOFT reflood behavior using the RELAP4/MOD6 code

International Nuclear Information System (INIS)

Lin, J.C.; Kee, E.J.; Grush, W.H.; White, J.R.

1978-01-01

The RELAP4/MOD6 computer code was used to predict the thermal-hydraulic transient for Loss-of-Fluid Test (LOFT) Loss-of-Coolant Accident (LOCA) experiments L2-2, L2-3, and L2-4. This analysis will aid in the development and assessment of analytical models used to analyze the LOCA performance of commercial power reactors. Prior to performing experiments in the LOFT facility, the experiments are modeled in counterpart tests performed in the nonnuclear Semiscale MOD 1 facility. A comparison of the analytical results with Semiscale data will verify the analytical capability of the RELAP4 code to predict the thermal-hydraulic behavior of the Semiscale LOFT counterpart tests. The analytical model and the results of analyses for the reflood portion of the LOFT LOCA experiments are described. These results are compared with the data from Semiscale

SAPHIRE technical reference manual: IRRAS/SARA Version 4.0

International Nuclear Information System (INIS)

Russell, K.D.; Atwood, C.L.; Sattison, M.B.; Rasmuson, D.M.

1993-01-01

This report provides information on the principles used in the construction and operation of Version 4.0 of the Integrated Reliability and Risk Analysis System (IRRAS) and the System Analysis and Risk Assessment (SARA) system. It summarizes the fundamental mathematical concepts of sets and logic, fault trees, and probability. The report then describes the algorithms that these programs use to construct a fault tree and to obtain the minimal cut sets. It gives the formulas used to obtain the probability of the top event from the minimal cut sets, and the formulas for probabilities that are appropriate under various assumptions concerning repairability and mission time. It defines the measures of basic event importance that these programs can calculate. The report gives an overview of uncertainty analysis using simple Monte Carlo sampling or Latin Hypercube sampling, and states the algorithms used by these programs to generate random basic event probabilities from various distributions. Further references are given, and a detailed example of the reduction and quantification of a simple fault tree is provided in an appendix

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.)

Evaluation report on CCTF CORE-I REFLOOD TEST Cl-1, (Run 010)

International Nuclear Information System (INIS)

Sudoh, Takashi; Murao, Yoshio.

1983-09-01

This report describes the effects of the loop flow resistance on the thermohydraulic behavior in the primary system during the reflood phase. The investigation is based on the results of the test Cl-1 which was performed with increased loop flow resistance in the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute. The loop flow resistance was about 40% higher in the present test than in the reference test Cl-5. The results of two tests were compared and the following conclusions were obtained: 1) The total loop flow rate and the core flooding rate were reduced by about 20% with the increased loop flow resistance 2) The core heat transfer was also lowered, then, the turnaround and the quench times extended at the locations above the core midplane. 3) The measured maximum temperature in the core was 50 K higher for the present test than for the reference test. (author)

Reflood Heat Transfer in SiC and Graphene Oxide Coated Tube

International Nuclear Information System (INIS)

Kim, Kyung Mo; Lee, Seung Won; Bang, In Cheol

2013-01-01

The reflood tests have been performed flowing water into bare tube and nanoparticles coated tube at constant flow rate (3 cm/s). The quenching curves have been obtained at atmospheric pressure. Finally, Scanning Electron Microscopy (SEM) images are acquired and contact angles are measured in order to observe the surface structures and wettability effect on cooling performance. The quenching time decreases and quenching velocity increases as the coating time of nanoparticles on the tube increases, because the nanoparticles deposited on the tube destabilize and rupture the vapor film early in the effect of increased Leidenfrost point temperature. The SiC nanoparticles coated tubes have better quenching performance than GO nanoparticles coated tubes. The SEM images and contact angle observations proved the enhanced wettability and rough surface due to deposition of SiC nanoparticles. And the wettability of GO nanoparticles coated tubes shows the increase at 600 s coating. But, the wettability decreases on GO nanoparticles tube coated for 900 s despite the enhanced quenching performance. Thus, the porous structure affects to the better cooling performance in case of GO nanoparticles coated tubes

SARAS 2: a spectral radiometer for probing cosmic dawn and the epoch of reionization through detection of the global 21-cm signal

Science.gov (United States)

Singh, Saurabh; Subrahmanyan, Ravi; Shankar, N. Udaya; Rao, Mayuri Sathyanarayana; Girish, B. S.; Raghunathan, A.; Somashekar, R.; Srivani, K. S.

2018-04-01

The global 21-cm signal from Cosmic Dawn (CD) and the Epoch of Reionization (EoR), at redshifts z ˜ 6-30, probes the nature of first sources of radiation as well as physics of the Inter-Galactic Medium (IGM). Given that the signal is predicted to be extremely weak, of wide fractional bandwidth, and lies in a frequency range that is dominated by Galactic and Extragalactic foregrounds as well as Radio Frequency Interference, detection of the signal is a daunting task. Critical to the experiment is the manner in which the sky signal is represented through the instrument. It is of utmost importance to design a system whose spectral bandpass and additive spurious signals can be well calibrated and any calibration residual does not mimic the signal. Shaped Antenna measurement of the background RAdio Spectrum (SARAS) is an ongoing experiment that aims to detect the global 21-cm signal. Here we present the design philosophy of the SARAS 2 system and discuss its performance and limitations based on laboratory and field measurements. Laboratory tests with the antenna replaced with a variety of terminations, including a network model for the antenna impedance, show that the gain calibration and modeling of internal additive signals leave no residuals with Fourier amplitudes exceeding 2 mK, or residual Gaussians of 25 MHz width with amplitudes exceeding 2 mK. Thus, even accounting for reflection and radiation efficiency losses in the antenna, the SARAS 2 system is capable of detection of complex 21-cm profiles at the level predicted by currently favoured models for thermal baryon evolution.

Escribiendo el silencio: la contemplación poética de Sara Pujol

Directory of Open Access Journals (Sweden)

Gala, Candelas

2005-06-01

Full Text Available Reading Sara Pujol Russell's El fuego tiende su aire (1999 and Intacto asombro de la luz del silencio (2001, is entering into a poetic world whose content and form are both innovative and complex. This poetry invites the reader into a state of meditative contemplation that seeks correspondences among different elements in reality as the key to access a superior type of knowledge. Pujol's writing moves in the fringes between voice and silence, art and nature, meaning and nothingness, identity and difference, it seeks to surpass its own verbal texture while only in that texture does it find articulation and only through it does transcendence becones accessible. The focus of the present reading centers on the point where Pujol's language seeks to transcend the disjunction between sign and surrounding, the point where synthesis is fusion about to dissolve.La lectura de El fuego tiende su aire (1999 e Intacto asombro en la luz del silencio (2001 de Sara Pujol Russell, supone la entrada en un discurso poético innovador y complejo tanto en contenido como en forma. Esta poesía invita a una contemplación meditativa que busca las correspondencias entre los diversos elementos de la realidad como clave para acceder a un conocimiento superior. Su escritura se mueve en los bordes entre la voz y el silencio, el arte y la naturaleza, el sentido y el vacío, la identidad y la diferencia, buscando trascender lo que está más allá de su misma urdimbre verbal pero que sólo en ella se configura, y sólo desde ella se puede acceder. En el movimiento de esta escritura por superar la disyunción entre signo y entorno, en esa síntesis a punto de disolverse, es donde se enfoca la lectura en este ensayo.

Analytical modeling of heat transfer during the reflooding phase of the LOCA: the UCFLOOD code

International Nuclear Information System (INIS)

Yadigaroglu, G.; Arrieta, L.A.

1980-01-01

A mechanistic model of bottom-reflooding heat transfer is described. From the hydrodynamic point of view the flow channel is divided into a single-phase liquid region, a continuous-liquid two-phase region, and a dispersed-liquid region. The void fraction is obtained from drift flux models. The onset of liquid entrainment is determined using a criterion based on the instability of the liquid core in the inverted-annular flow regime. For heat transfer calculations, the channel is also divided into a number of regions. The heat transfer coefficients are functions of the local flow conditions. Quench front propagation is treated separately by a model including the effects of axial conduction. Good agreement of calculated and experimental results has been obtained

Evaluation report on CCTF core-I reflood tests Cl-17(Run 36) and Cl-20(Run 39)

International Nuclear Information System (INIS)

Murao, Yoshio; Iguchi, Tadashi

1983-02-01

In the safety analysis of the reflood phase of a PWR LOCA, the core thermo-hydrodynamic behavior is analyzed as a phenomena in a single channel core. In other words, the core thermo-hydrodynamic behavior is treated one-dimensionally. In order to confirm the validity of the one-dimensional treatment, tests named Asymmetric power test Cl-17(Run 36) and Asymmetric temperature test Cl-20(Run 39) were performed, whose test conditions were similar to the base case test Cl-5(Run 14) (the reference test for the parametric effect tests) except for the power distribution and the initial temperature distribution in core, respectively. First the results of the base case test were investigated. And the results of Asymmetric power test and Asymmetric temperature test were compared with the results of the base case test. The main conclusions are as follows: (1) The water accumulation was observed along the whole core even above quench front almost simultaneously just after reflood initiation and almost terminated within 20 to 30 seconds. The flow pattern was recognized as a slug flow. (2) In the lower two-thirds of the core, bottom quench was observed and the core water accumulation was one-dimensional even under the thermally asymmetric conditions. (3) In the upper portion of the core, the multi-dimensional effect was observed, i.e. the top quench occurred locally. (4) The water accumulation behavior or void fraction in the lower two-thirds of the core, i.e. our concerning region for the peak clad temperature analysis, and the core behavior for system analysis can be one-dimensionally analyzed with a representing single channel core. (author)

Os suínos da raça Bísara – oportunidades e desafios

OpenAIRE

Carvalho, Marieta

2015-01-01

A procura mundial de produtos de origem animal aumentará cerca de 70% em 2050. Estima-se que mil milhões de pobres dependam dos animais para a sua alimentação e criação de riqueza1. A carne de porco é um dos alimentos mais consumidos mundialmente, representando em 2012: 43,3% em todo o mundo2, 45,9% na União Europeia1 e 39,8 % em Portugal da carne total consumida3. Este trabalho tem como objetivo, fazer uma caracterização atual da suinicultura, com base nos suínos da raça Bísara, enumer...

Entrainment of droplets during the reflood phase of a LOCA and the influence of the channel geometry

International Nuclear Information System (INIS)

van der Molen, S.B.; Galjee, F.W.B.M.

1979-01-01

This paper describes the different reflood phenomena and the most relevant appearing two phase flow patterns such as the inverted annular, the annular and dispersed flow regimes and the transition between the different regimes. From high speed films it is clear that the onset of entrainment in case of the liquid column type quench mode occurs simultaneously with the transition of churn flow to a dispersed flow whereas in case of the dryout type quench mode with an annular flow at the quench front, the onset of entrainment is caused by the interaction of the vapor flow on the liquid film flow. 14 refs

Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

International Nuclear Information System (INIS)

Murao, Y.; Iguchi, T.; Sugimoto, J.

1988-09-01

The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

Nuclear spectroscopy of exotic nuclei with the Sara/Igisol facility

International Nuclear Information System (INIS)

Beraud, R.; Emsallem, A.; Astier, A.; Duffait, R.; Le Coz, Y.; Redon, N.; Barneoud, D.; Genevey, J.; Gizon, A.

1994-11-01

The authors review their recent studies on alpha and beta decay of exotic nuclei performed with the on-line mass separator at the Igisol/Sara facility in Grenoble. The experiments using charged particle induced fission have given new information on production cross section and properties of n-rich nuclei with A=110-130 whereas by means of heavy ion induced fusion evaporation reactions the authors have investigated two regions close to the proton drip line around A=180 and A=130. This paper gives first a brief description of the Igisol technique and shows its application in case of two different production modes: charged particle-induced fission and heavy ion -induced fusion-evaporation reactions. The systematic study of the low-lying levels in n-rich Ru isotopes has allowed to show an axial symmetry breaking, whereas complementary investigations are necessary to clarify the case of 180 Tl decay. A number of new spectroscopic data such as new isotopes identification have been gained in the region of light rare earth nuclei. (N.T.)

ASTEC code development, validation and applications for severe accident management within the CESAM European project - 15392

International Nuclear Information System (INIS)

Van Dorsselaere, J.P.; Chatelard, P.; Chevalier-Jabet, K.; Nowack, H.; Herranz, L.E.; Pascal, G.; Sanchez-Espinoza, V.H.

2015-01-01

ASTEC, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European research on the domain. The CESAM project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. Within the CESAM project 3 main types of research activities are performed: -) further validation of ASTEC models important for SAM, in particular for the phenomena being of importance in the Fukushima-Daichi accidents, such as reflooding of degraded cores, pool scrubbing, hydrogen combustion, or spent fuel pools behaviour; -) modelling improvements, especially for BWR or based on the feedback of validation tasks; and -) ASTEC applications to severe accident scenarios in European NPPs in order to assess prevention and mitigation measures. An important step will be reached with the next major ASTEC V2.1 version planned to be delivered in the first part of 2015. Its main improvements will concern the possibility to simulate in details the core degradation of BWR and PHWR and a model of reflooding of severely degraded cores. A new user-friendly Graphic User Interface will be available for plant analyses

Modeling of quench front progression and heat transfer by radiation during reflooding of a tubular test section

International Nuclear Information System (INIS)

Clement, P.; Deruaz, R.

1976-01-01

Heat transfer modeling is presented in the scope of emergency core cooling. The rewetting of a hot dry wall during reflooding is a conduction-controlled phenomenon described by a model of heat-transfer coefficient. Upstream of the quench front, a two-dimensional approach involving both axial and transverse (or radial) heat conduction is discussed in view of thick walls, high quench front velocities and nucleate boiling. Downstream of the quench-front, high wall temperatures are reached so that a thermal radiation model is required to separate the different mechanisms of heat transfer. An attempt is made to consider radiation between walls, water droplets and vapor, with scattering emission and absorption of the two phases

Pushing and Pulling Sara: A Case Study of the Contrasting Influences of High School and University Experiences on Engineering Agency, Identity, and Participation

Science.gov (United States)

Godwin, Allison; Potvin, Geoff

2017-01-01

This manuscript reports a longitudinal case study of how one woman, Sara, who had previously considered dropping out of high school, authored strong mathematics and science identities and purposefully exhibited agency through her experiences in high school science. These experiences empowered her to choose an engineering major in college; however,…

Post-BEMUSE Reflood Model Input Uncertainty Methods (PREMIUM) Benchmark Phase II: Identification of Influential Parameters

International Nuclear Information System (INIS)

Kovtonyuk, A.; Petruzzi, A.; D'Auria, F.

2015-01-01

The objective of the Post-BEMUSE Reflood Model Input Uncertainty Methods (PREMIUM) benchmark is to progress on the issue of the quantification of the uncertainty of the physical models in system thermal-hydraulic codes by considering a concrete case: the physical models involved in the prediction of core reflooding. The PREMIUM benchmark consists of five phases. This report presents the results of Phase II dedicated to the identification of the uncertain code parameters associated with physical models used in the simulation of reflooding conditions. This identification is made on the basis of the Test 216 of the FEBA/SEFLEX programme according to the following steps: - identification of influential phenomena; - identification of the associated physical models and parameters, depending on the used code; - quantification of the variation range of identified input parameters through a series of sensitivity calculations. A procedure for the identification of potentially influential code input parameters has been set up in the Specifications of Phase II of PREMIUM benchmark. A set of quantitative criteria has been as well proposed for the identification of influential IP and their respective variation range. Thirteen participating organisations, using 8 different codes (7 system thermal-hydraulic codes and 1 sub-channel module of a system thermal-hydraulic code) submitted Phase II results. The base case calculations show spread in predicted cladding temperatures and quench front propagation that has been characterized. All the participants, except one, predict a too fast quench front progression. Besides, the cladding temperature time trends obtained by almost all the participants show oscillatory behaviour which may have numeric origins. Adopted criteria for identification of influential input parameters differ between the participants: some organisations used the set of criteria proposed in Specifications 'as is', some modified the quantitative thresholds

A moving subgrid model for simulation of reflood heat transfer

International Nuclear Information System (INIS)

Frepoli, Cesare; Mahaffy, John H.; Hochreiter, Lawrence E.

2003-01-01

In the quench front and froth region the thermal-hydraulic parameters experience a sharp axial variation. The heat transfer regime changes from single-phase liquid, to nucleate boiling, to transition boiling and finally to film boiling in a small axial distance. One of the major limitations of all the current best-estimate codes is that a relatively coarse mesh is used to solve the complex fluid flow and heat transfer problem in proximity of the quench front during reflood. The use of a fine axial mesh for the entire core becomes prohibitive because of the large computational costs involved. Moreover, as the mesh size decreases, the standard numerical methods based on a semi-implicit scheme, tend to become unstable. A subgrid model was developed to resolve the complex thermal-hydraulic problem at the quench front and froth region. This model is a Fine Hydraulic Moving Grid (FHMG) that overlies a coarse Eulerian mesh in the proximity of the quench front and froth region. The fine mesh moves in the core and follows the quench front as it advances in the core while the rods cool and quench. The FHMG software package was developed and implemented into the COBRA-TF computer code. This paper presents the model and discusses preliminary results obtained with the COBRA-TF/FHMG computer code

Reading Sara Pujol Russell’s Poetry of Contemplation and Connection

Directory of Open Access Journals (Sweden)

Anita M. Hart

2012-06-01

Full Text Available Sara Pujol Russell’s poetry captures a process of expanding consciousness and personal renewal. Through contemplation and attention to nature, the poet-speaker in her works generates a sense of connection that moves her beyond daily concerns. Pujol’s poetry is both metaphysical and also different in that it resists easy classification and is not representative of mainstream trends. This essay approaches the distinctiveness of Pujol’s work by studying selected poems from her third book of poetry in Spanish, Para decir sí a la carencia, sí a la naranja, al azafrán en el pan (2004 ‘To Say Yes to Lack, Yes to the Orange, to the Saffron in the Bread.’ Incorporating philosopher María Zambrano’s thoughts on contemplation, it shows Pujol’s poet-speaker establishing a connection with nature and spirit, experiencing a heightened consciousness, and searching for expression. The poetic language is characterized by vision, intimacy, enigma, and contradiction. In its subjective, intuitive way, Pujol’s work reveals the poet-speaker’s winding path of discovery and challenges the reader to look closely inside and outside in engaging life’s mysteries.

Women Empowerment in the Realms of Institutionalized Religion and Patriarchy: El Saadawi’s Firdaus and Yezierska’s Sara as Examples

OpenAIRE

Abdullah K. Shehabat

2016-01-01

This paper explains how the two protagonists, Firdaus and Sara, successfully paved their own ways in search of self-liberation despite the authoritarian patriarchy and institutionalized religions that plagued them. El Saadawi's Woman at Point Zero and Yezierska’s Bread Givers represent the fruitful struggle these protagonists experienced as they come to forge an identity and be themselves. The paper argues that the protagonists manage to free themselves, establish their own spiritual homes at...

Measurement of grid spacer's enhanced droplet cooling under reflood condition in a PWR by LDA

International Nuclear Information System (INIS)

Lee, S.L.; Sheen, H.J.; Cho, S.K.; Issapour, I.; Hua, S.Q.

1984-01-01

Reported is an experiment designed for the measurements of grid spacer's enhanced droplet cooling under reflood condition at elevated temperatures in a steam environment. The flow channel consists of a simulated 1.60m-long pressurized water reactor (PWR) fuel rod bundle of 2 x 2 electrically heated rods. Embedded thermocouples are used to measure the rod cladding temperature at various axial levels and an unshielded Chromel-Alumel thermocouple sheathed by a small Inconel tube is traversed in the center of the subchannel to measure the temperatures of the water and steam coolant phases at various levels. The droplet dynamics across the grid spacer is directly obtained by a special laser-Doppler anemometry technique for the in situ simultaneous measurement of velocity and size of droplets through two observation windows on the test channel, one immediately before and one immediately after the grid spacer. Some results are presented and analyzed

Reentrainment of droplet from grid spacer in mist flow portion of LOCA reflood of PWR

International Nuclear Information System (INIS)

Lee, S.L.; Cho, S.K.; Sheen, H.J.

1983-01-01

An investigation is made on the influence of a quenched grid spacer on the greatly enhanced heat transfer from heated fuel rods during the mist flow phase of emergency reflood of loss of coolant accident (LOCA) of a pressurized water reactor (PWR). The situation for the case of a dry grid spacer before its quenching has not been covered in this study. The experimental technique used is a relatively simple optical scheme which combines the reference-mode laser-Doppler anemometry making use of the scattering of a light beam from a droplet. The results reveal that the large droplets in the mist flow, which are intercepted by the grid spacer, are responsible for the creation of a large number of smaller droplets. These small droplets, due to their large surface area to mass ratios, can serve as superb evaporative cooling agents to heat transfer downstream of the grid spacer

The comparison of naturally weathered oil and artificially photo-degraded oil at the molecular level by a combination of SARA fractionation and FT-ICR MS

International Nuclear Information System (INIS)

Islam, Ananna; Cho, Yunju; Yim, Un Hyuk; Shim, Won Joon; Kim, Young Hwan; Kim, Sunghwan

2013-01-01

Highlights: • Weathered oils from the Hebei Spirit oil spill and photo degraded oils are compared. • We investigate changes of polar species at the molecular level by 15T FT-ICR MS. • Significant reduction of sulfur class compounds in saturates fraction is observed. • The relative abundance of protonated compounds (presumably basic nitrogen compounds) increase after degradation. • Changes of polar compounds occurred by natural and photo degradation are similar. -- Abstract: Two sets of oil samples, one obtained from different weathering stages of the M/V Hebei Spirit oil spill site and the other prepared by an in vitro photo-degradation experiment, were analyzed and compared at the molecular level by atmospheric pressure photo-ionization coupled with Fourier transform ion cyclotron resonance mass spectrometry (FT-ICR MS). For a more detailed comparison at the molecular level, the oil samples were separated into saturate, aromatic, resin, and asphaltene (SARA) fractions before MS analysis. Gravimetric analysis of the SARA fractions revealed a decreased weight percentage of the aromatic fraction and an increased resin fraction in both sets of samples. Molecular-level investigations of the SARA fractions showed a significant reduction in the S 1 class in the saturate fraction and increase of S 1 O 1 class compounds with high DBE values in resin fraction. Levels of N 1 and N 1 O 1 class compounds resulting in protonated ions (presumably basic nitrogen compounds) increased after degradation compared to compounds generating molecular ions (presumably non-basic nitrogen compounds). This study revealed changes occurring in heteroatom polar species of crude oils such as sulfur and nitrogen containing compounds that have not been easily detected with conventional GC based techniques

Large scale reflood test with cylindrical core test facility (CCTF). Core I. FY 1979 tests

International Nuclear Information System (INIS)

Murao, Yoshio; Akimoto, Hajime; Okubo, Tsutomu; Sudoh, Takashi; Hirano, Kenmei

1982-03-01

This report presents the results of analysis of the data obtained in the CCTF Core I test series (19 tests) in FY. 1979 as an interim report. The Analysis of the test results showed that: (1) The present safety evaluation model on the reflood phenomena during LOCA conservatively represents the phenomena observed in the tests except for the downcomer thermohydrodynamic behavior. (2) The downcomer liquid level rose slowly and it took long time for the water to reach a terminal level or the spill-over level. It was presume that such a results was due to an overly conservative selection of the ECC flow rate. This presumption will be checked against a future test result for an increased flow rate. The loop-seal-water filling test was unsuccessful due to a premature power shutdown by the core protection circuit. The test will be conducted again. The tests to be performed in the future are summerized. Tests for investigation of the refill phenomena were also proposed. (author)

sarA negatively regulates Staphylococcus epidermidis biofilm formation by modulating expression of 1 MDa extracellular matrix binding protein and autolysis‐dependent release of eDNA

DEFF Research Database (Denmark)

Christner, Martin; Heinze, Constanze; Busch, Michael

2012-01-01

to biofilm formation in mutant 1585ΔsarA. Increased eDNA amounts indirectly resulted from upregulation of metalloprotease SepA, leading to boosted processing of autolysin AtlE, in turn inducing augmented autolysis and release of eDNA. Hence, this study identifies sarA as a negative regulator of Embp‐ and e...

FLECHT-SEASET 21-rod bundle flow blockage heat transfer during reflood

International Nuclear Information System (INIS)

Loftus, M.; Hochreiter, L.; Lee, N.

1983-01-01

The effect of various flow blockage shapes and distributions during a PWR reflood was investigated using six 21-rod bundles with full length, internally heated, cosine power-shaped electrical rods. The flow blockage shapes, simulating the fuel rod clad ballooning, were made of thin-wall stainless steel tubes hydroformed into a short, concentric shape and along, nonconcentric shape. The blockage sleeves were distributed both coplanar, with all sleeves located at the same elevation, and non-coplanar. The initial and boundary conditions were varied to include parametric effects of pressure, inlet water temperature, and primarily, flooding rate. The initial mid-plane rod temperature was 871 0 C (1600 0 F) in all tests. Rod and vapor temperature measurements were made throughout the rod bundle with emphasis on the blockage region. The rod heat transfer downstream of the blockage was found to be greater for rods in a blocked bundle than for similar rods in an unblocked bundle. The heat transfer improvement decreases both with time after flood initiation and as the distance increased downstream of the blockage. The improvement in the heat transfer is attributed primarily to the breakup of the water droplets entrained in the steam flow. The smaller droplets subsequently evaporate and desuperheat the steam, which then improves the heat transfer between the rods and the steam in and downstream of the blockage zone

TRADE instructional materials for SARA/OSHA training. Volume 2, Managers and supervisors training

Energy Technology Data Exchange (ETDEWEB)

1989-03-01

This document provides instructional materials for an eight-hour training course for managers and supervisors of hazardous waste sites. It is one of three volumes of course materials TRADE is preparing to help DOE contractor training staff comply with 29 CFR 1910.120, the Occupational Health and Safety Administration (OSHA) rule that implements Title I of the Superfund Amendments and Reauthorization Act (SARA) of 1986. OSHA`s final rule for hazardous waste operators was published in the Federal Register of March 6, 1989 (54 FR 9294). Combined with the materials in Volumes I and III and with appropriate site-specific information, these materials will help DOE contractors to meet the requirements of 1910.120 (e) that ``on-site management and supervisors directly responsible for, or who supervise employees engaged in, hazardous waste operations`` receive the same initial training as that of the employees they supervise and at least eight additional hours of specialized training in managing hazardous waste operations.

Assessment of current safety evaluation analysis on reflood behavior during PWR-LOCA by using CCTF data

International Nuclear Information System (INIS)

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

1987-01-01

This paper describes an assessment result on conservatism of current safety analysis concerning reflood behavior during a LOCA in a PWR by using the experimental data with cylindrical core test facility (CCTF) performed at Japan Atomic Energy Research Institute (JAERI). WREM code is selected for a representative of current safety analyses. The predicted peak clad temperature with the WREM code was higher than the data, and it was confirmed that the WREM code had the overall conservatism against CCTF data. The WREM code predicted the reasonable core boundary conditions and it was found that the conservatism of the code came mainly from the calculations on the incore thermal hydraulics and clad temperature. In addition, it was found that the conservatism of the WREM code against the CCTF data could be attributed to the neglection of horizontal fluid mixing between subchannels, the neglection of the heat transfer enhancement due to the radial core power profile, and the usage of the heat transfer correlations conservative against CCTF data. (author)

Effects of fuel relocation on reflood in a partially-blocked rod bundle

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Jae [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 34134 (Korea, Republic of); Kim, Jongrok; Kim, Kihwan; Bae, Sung Won [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of)

2017-02-15

Ballooning of the fuel rods has been an important issue, since it can influence the coolability of the rod bundle in a large-break loss-of-coolant accident (LBLOCA). Numerous past studies have investigated the effect of blockage geometry on the heat transfer in a partially blocked rod bundle. However, they did not consider the occurrence of fuel relocation and the corresponding effect on two-phase heat transfer. Some fragmented fuel particles located above the ballooned region may drop into the enlarged volume of the balloon. Accordingly, the fuel relocation brings in a local power increase in the ballooned region. The present study’s objective is to investigate the effect of the fuel relocation on the reflood under a LBLOCA condition. Toward this end, experiments were performed in a 5 × 5 partially-blocked rod bundle. Two power profiles were tested: one is a typical cosine shape and the other is the modified shape considering the effect of the fuel relocation. For a typical power shape, the peak temperature in the ballooned rods was lower than that in the intact rods. On the other hand, for the modified power shape, the peak temperature in the ballooned rods was higher than that in the intact rods. Numerical simulations were also performed using the MARS code. The tendencies of the peak clad temperatures were well predicted.

Experiment of the downcomer effective water head during a reflood phase of PWR LOCA

International Nuclear Information System (INIS)

Sudo, Yukio; Murao, Yoshio

1978-12-01

The results and analysis are described of a downcomer effective water head experiment. Downcomer effective water head is the driving force to feed an emergency coolant to the core during a reflood phase of PWR LOCA. The test rig has dimensions of the full-scale height and gap. Experimental conditions are: downcomer wall temperature = 250 0 -- 300 0 C, back pressure = 1 atm, coolant temperature = 98 0 -- 100 0 C, extraction water velocity = 0 -- 2 cm/s, and gap size = 200 mm. The effective water head histories obtained by experiment were compared with those predicted from the heat release from the downcomer walls. The heat release was calculated from the temperature histories indicated by thermocouples instrumented in and on the walls during experiment. The following were revealed: (1) The relation of heat flux and superheat (q vs ΔT sub(s)) obtained in the experiment is much different from that in pool boiling. (2) The predicted effective water head is in good agreement with the experimental one after 120 sec from the initiation of coolant injection. (3) The effect of extraction water velocity is negligible. (4) The effect of initial wall temperatures is evident. (author)

The acquisition and supervision system of S.A.R.A.'s (Accelerator system Rhone-Alpes) parameters

International Nuclear Information System (INIS)

Iazzourene, F.

1982-01-01

The acquisition and supervision system of SARA's (Systeme Accelerateur Rhone-Alpes) parameters is built up. The basic hardware consists of: - A PDP 11/10 computer with a 64 K bytes memory capacity. The system and load device is a floppy disk of 28 megabytes capacity. - A CAMAC crate including a data logger with 224 input channels, a terminal driver (JTY21) and three modules designed for reading out a few digital data, for instance polarities of power supplies. The software provides three distinct programs: AKITS, which uses 3 commands, detects and signals functioning defects in the CAMAC modules used. AKIDO which uses 11 commands, is the acquisition and organization program of the accelerator's functioning parameters. AKISUR is the supervision program of the functioning parameter's stability, within a fixed gap, during the accelerator running [fr

Instructional materials for SARA/OSHA training. Volume 1, General site working training

Energy Technology Data Exchange (ETDEWEB)

Copenhaver, E.D.; White, D.A.; Wells, S.M. [Oak Ridge National Lab., TN (United States)

1988-04-01

This proposed 24 hour ORNL SARA/OSHA training curriculum emphasizes health and safety concerns in hazardous waste operations as well as methods of worker protection. Consistent with guidelines for hazardous waste site activities developed jointly by National Institute for Occupational Safety and Health, Occupational Safety and Health Administration, US Coast Guard, and the Envirorunental Protection Agency, the program material will address Basic Training for General Site Workers to include: ORNL Site Safety Documentation, Safe Work Practices, Nature of Anticipated Hazards, Handling Emergencies and Self-Rescue, Employee Rights and Responsibilities, Demonstration of Use, Care, and Limitations of Personal Protective, Clothing and Equipment, and Demonstration of Monitoring Equipment and Sampling Techniques. The basic training courses includes major fundamentals of industrial hygiene presented to the workers in a format that encourages them to assume responsibility for their own safety and health protection. Basic course development has focused on the special needs of ORNL facilities. Because ORNL generates chemical wastes, radioactive wastes, and mixed wastes, we have added significant modules on radiation protection in general, as well as modules on radiation toxicology and on radiation protective clothing and equipment.

Targets and special materials

International Nuclear Information System (INIS)

Blanc, R.; Bouriant, M.; Richaud, J.P.

1997-01-01

The target preparation group supplied a large number of samples to nuclear physicists for experiments using SARA and also other accelerators throughout the world. Particular preparation and projects include: 208 Pb, 116 Cd, 6 LiF, 123 Sb, In and Ta targets, strippers for SARA and GANIL, optical silicone disks for POLDER and GRAAL experiments, active participations for the AMS project and finally filament preparation for the GENEPI project. (authors)

Feminism and Faith: Exploring Christian Spaces in the Writing of Sara Maitland and Michèle Roberts

Directory of Open Access Journals (Sweden)

Arina LUNGU-CIRSTEA

2011-03-01

Full Text Available En 1983, les féministes britanniques Sara Maitland et Jo Garcia ont publié Walking on the Water (London : Virago, une collection d’essais, de récits, de poèmes et de photos produits par des femmes sur le thème de la spiritualité. Les contributrices ont été en particulier invitées à explorer la relation entre leur identité féministe et leurs croyances religieuses. Le ton de ces contributions varie fortement, allant de l’envie passionnée de concilier les objectifs du féminisme avec le christianisme à un rejet total de l’Eglise comme institution patriarcale suprême. Cet article met en dialogue des récits diamétralement opposés du rapport entre christianisme et féminisme en s’intéressante plus particulièrement à deux des contributrices, Sara Maitland (1950 - et Michèle Roberts (1949 - . Ces deux écrivaines, qui se sont activement impliquées dans les mouvements féministes des années 1970, ont toutes deux lutté pour se réconcilier avec leur héritage chrétien. Néanmoins, alors que Maitland tente essentiellement de revisiter le christianisme en y incorporant les points essentiels d’une idéologie féministe, Roberts sent le besoin impérieux de se défaire de son identité religieuse afin de devenir indépendante ; en effet, dans son autobiographie Paper Houses (2007 elle décrit son éducation catholique comme “autoritaire et misogyne” (16. Cet article explore les façons dont l’identité spirituelle se construit dans le jeu complexe des interactions entre féminisme et foi. Il se propose, dans une perspective comparatiste, d’analyser d’une part le recueil de nouvelles de Sara Maitland intitulé A Book of Spells, et d’autre part, le roman acclamé de Michèle Roberts, Daughters of the House. Dans ces écrits, Maitland et Roberts ont un objectif commun qui est de renégocier la place des femmes dans l’histoire chrétienne dont elles reconnaissent – il est vrai à partir de perspectives diff

Two-phase flow parameters of a downcomer boiling during a postulated reflood phase of APR1400; Parametres d'ecoulement diphasique dans un ebullition a la cuve en acier pendant un refroidissement d'un APRP de APR1400

Energy Technology Data Exchange (ETDEWEB)

Byong-Jo, Yun; Dong-Jin, Euh; Chul-Hwa, Song [Korea Atomic Energy Research Inst. (Korea, Republic of)

2009-07-01

Downcomer boiling phenomena is one of the key issues for a postulated large-break LOCA (LBLOCA) in a conventional pressurized water reactor, because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling. To investigate the thermal hydraulic behavior in the downcomer region of the APR1400, a test program for a downcomer boiling is being progressed for the reflood phase of a postulated LBLOCA. Test was performed in a one side heated rectangular test channel which was designed by adopting a full-pressure, full-height, and full-size downcomer-gap approach, but with the circumferential length was reduced 47.08-fold. The test consists of two steps: (I) for the global two-phase flow parameters and (II) for the local two-phase flow parameters. The step-I test has already been completed. In the present paper, the experimental results of the step-II test are introduced. (authors)

Memoria de la guerra civil española. Entre el sol y la tormenta de Sara Berenguer

Directory of Open Access Journals (Sweden)

Helena López

2013-07-01

Full Text Available En este artículo pretendo analizar dos cuestiones en relación con las memorias de guerra de la militante anarquista Sara Berenguer. En primer lugar, quiero atender a las estrategias discursivas (clase, género, sexualidad de construcción de la subjetividad en este texto. Además, me interesa indagar cómo estos discursos se interseccionan con la posición espacio-temporal (tiempo de la memoria y tiempo-espacio del exilio del sujeto autobiográfico. Mi objetivo principal es proponer una problematización de los conceptos de “memoria colectiva” y de “experiencia femenina” y reivindicar, por lo tanto, la relevancia tanto teórica como política de análisis críticos basados en narrativas personales.

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II forced feed reflood tests

International Nuclear Information System (INIS)

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

1987-01-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase of a PWR-LOCA. It was revealed in the previous Slab Core Test Facility (SCTF) Core-II test results that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. In order to separately evaluate the effect of the radial power (Q) distribution itself and the effect of the radial temperature (T) distribution, four tests were performed with steep Q and T, flat Q and T, steep Q and flat T, and flat Q and steep T. Based on the test results, it was concluded that the radial temperature distribution which accompanied the radial power distribution was the dominant factor of the two-dimensional thermal-hydraulic behavior in the core during the initial period. Selected data from these four tests are also presented in this report. Some data from Test S2-12 (steep Q, T) were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

Direct ECC bypass phenomena in the MIDAS test facility during LBLOCA reflood phase

International Nuclear Information System (INIS)

Yun, B. J.; Kweon, T. S.; Ah, D. J.; Ju, I. C.; Song, C. H.; Park, J. K.

2001-01-01

This paper describes the experimental results of ECC Direct Bypass Phenomena in the downcomer during the late reflood phase of LBLOCA of the reactor that adopts Direct Vessel Injection as a ECC system. The experiments have been performed in MIDAS test facility using superheated steam and water. The test condition was determined, based on the preliminary analysis of TRAC code, from modified linear scaling method of 1/4.93 length scale. To measure the direct bypass fraction according to the nozzle location, separate effect tests have been performed in case of DVI-4(farthest from broken cold leg) injection, DVI-2(closest to broken cold leg) injection, and DVI-2 and 4 injection, respectively. Also the test was carried out varying the steam flow rate greatly to investigate the effect of steam flow rate on the direct bypass fraction of ECC water. Test results show that the direct bypass fraction of ECC water depends significantly on the injected steam mass flow rate. DVI-4 tests show that the direct bypass fraction increases drastically as the steam flow rate increases. However, in DVI-2 test most of the injected ECC water penetrates into lower downcomer. The direct bypass characteristic in each of DVI-2 and DVI-4 tests is reflected into the direct bypass characteristic curve of DVI-2 and 4 test. The steam condensation reaches to the theoretically allowable maximum value

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)

A dinâmica dos mercados associados ao suínos de raça Bísara na Terra Fria Transmonta

OpenAIRE

Carvalho, Marieta

2014-01-01

A procura mundial de produtos de origem animal aumentará cerca de 70% em 2050, com especial relevância para a de origem do porco (FAO, 2014). Este trabalho visa estudar: · A dinâmica dos mercados associados aos suínos da raça Bísara: animais, carne e fumeiro de Vinhais. · Os circuitos de comercialização, os preços ao criador de suínos de animais vivos, da carne e fumeiro de Vinhais. A metodologia utilizada baseou-se na análise descritiva, referente aos de 2013 e 2014, dos ...

Analysis of Seven NEPTUN-III (Tight-Lattice) Bottom-Flooding Experiments with RELAP5/MOD3.3/BETA

International Nuclear Information System (INIS)

Analytis, G.Th.

2004-01-01

Seven tight-lattice NEPTUN-III bottom-flooding experiments are analyzed by using the frozen version of RELAP5, RELAP5/MOD3.3/BETA. This work is part of the Paul Scherrer Institute (PSI) contribution to the High Performance Light Water Reactor (HPLWR) European Union project and aims at assessing the capabilities of the code to model the reflooding phenomena in a tight hexagonal lattice (which was one of the core geometries considered at the time for an HPLWR) following a hypothetical loss-of-coolant accident scenario. Even though the latest version of the code has as a default the new PSI reflood model developed by the author, which was tested and assessed against reflooding data obtained at standard light water reactor lattices, this work shows that for tight lattices, the code underpredicts the peak clad temperatures measured during a series of reflooding experiments performed at the NEPTUN-III tight-lattice heater rod bundle facility. The reasons for these differences are discussed, and the (possible) changes needed in the framework of RELAP5/MOD3.3 for improving the modeling of reflooding in tight lattices are investigated

Análise SWOT global da exploração dos suínos da raça Bísara

OpenAIRE

Carvalho, Marieta

2015-01-01

A procura mundial de produtos de origem animal aumentará cerca de 70% em 2050. Estima-se que mil milhões de pobres dependam dos animais para a sua alimentação e criação de riqueza (FAO, 2014). A raça Bísara, são suínos autóctones portugueses do tronco Celta em risco de extinção. Apesar do seu reduzido efetivo representa para as populações locais um elevado peso económico e social. Este trabalho tem como objetivo, fazer uma caracterização atual da suinicultura, com base nos suínos da ra...

Drop size measurements and entrainment in APR1400 during LBLOCA reflood phase

Energy Technology Data Exchange (ETDEWEB)

Lee, Eo Hwak

2010-02-15

A study has been performed to investigate droplet size in the nuclear reactor of APR1400 during LBLOCA reflood phase and to develop droplet entrainment and deposition models for SPACE (Safety and Performance CodE) which is a safety analysis tool for PWR being developed in Korea. A freezing technique for measuring the size of droplets was developed to obtain the droplet size distribution in horizontal annular flow in a pipe with a 37.1 mm diameter. Droplets are frozen by using an extremely low temperature nitrogen gas with liquid film extraction. They are then photographed with a microscope and a CCD camera and measured by means of an image process. The results are compared with various experimental data. The droplet sizes measured by the freezing technique are comparable with those measured by other methods at a high superficial air velocity (of 50 m/s). However, because of the film extraction problem, the droplet sizes measured at a low superficial air velocity of less than 40 m/s are higher than those measured by other methods. A present method suggested for predicting the Sauter mean diameter is based on the maximum droplet size correlation for the experimental data, with and without liquid film extraction. The average droplet size is remarkably smaller downstream of the liquid film extractor because large droplets from the liquid film are excluded. In order to understand and to predict a heat transfer between superheated steam and droplets properly during reflood phase of LBLOCA, it is very important to measure broken droplet sizes by spacer grids. A study, therefore, has been performed to investigate droplet size in rod bundles with spacer grids and to develop a spacer grid droplet breakup model for safety analysis codes. Experiments were conducted with liquid droplets (SMD of 300∼700 μm) and various spacer grids at superficial air velocity of 10 m/s and 20 m/s based on FLECHT SEASET. The test channel and the grids were heated to 150 .deg. C to prevent

Drop size measurements and entrainment in APR1400 during LBLOCA reflood phase

International Nuclear Information System (INIS)

Lee, Eo Hwak

2010-02-01

A study has been performed to investigate droplet size in the nuclear reactor of APR1400 during LBLOCA reflood phase and to develop droplet entrainment and deposition models for SPACE (Safety and Performance CodE) which is a safety analysis tool for PWR being developed in Korea. A freezing technique for measuring the size of droplets was developed to obtain the droplet size distribution in horizontal annular flow in a pipe with a 37.1 mm diameter. Droplets are frozen by using an extremely low temperature nitrogen gas with liquid film extraction. They are then photographed with a microscope and a CCD camera and measured by means of an image process. The results are compared with various experimental data. The droplet sizes measured by the freezing technique are comparable with those measured by other methods at a high superficial air velocity (of 50 m/s). However, because of the film extraction problem, the droplet sizes measured at a low superficial air velocity of less than 40 m/s are higher than those measured by other methods. A present method suggested for predicting the Sauter mean diameter is based on the maximum droplet size correlation for the experimental data, with and without liquid film extraction. The average droplet size is remarkably smaller downstream of the liquid film extractor because large droplets from the liquid film are excluded. In order to understand and to predict a heat transfer between superheated steam and droplets properly during reflood phase of LBLOCA, it is very important to measure broken droplet sizes by spacer grids. A study, therefore, has been performed to investigate droplet size in rod bundles with spacer grids and to develop a spacer grid droplet breakup model for safety analysis codes. Experiments were conducted with liquid droplets (SMD of 300∼700 μm) and various spacer grids at superficial air velocity of 10 m/s and 20 m/s based on FLECHT SEASET. The test channel and the grids were heated to 150 .deg. C to prevent

La difesa della donna ebrea: Sara Copio Sullam e Debora Ascarelli

Directory of Open Access Journals (Sweden)

Umberto Fortis

2014-04-01

Full Text Available Sara Copio Sullam and Debora Ascarelli, the best-known Jewish women poets of the age of the Italian ghetto, have often been studied with a focus on the distinctive features of their writing: the former as a translator of sacred texts, the latter as the author of original verses, sometimes written as a risposta per le rime (reply through rhymes or in defence of her orthodoxy. They share, however, a common theme, which has often been neglected: the defence of the Jewish woman, which is overt in some of Copio’s sonnets and prose, whereas it has not yet been properly pointed out in some of Ascarelli’s verses. This paper aims at bringing this theme to the fore not only in Copio, through a rereading of some passages of her Manifesto, but also in Debora Ascarelli, through an analysis of her few original verses. These hendecasyllables certainly reflect the Petrarchan and classicistic atmosphere of the late 16th century in Rome, but they are far from the mainstream modes of women’s poetry of the age. They allow us, therefore, to highlight, besides the differences, the similarities between the two poetesses; their commitment, in the late 16th and early 17th century, to the intellectual and moral defence of the Jewish woman, in a context of general depreciation of women, but also in the background of the claim to the “nobility and excellence of women”.

A study of the loss of coolant accident

International Nuclear Information System (INIS)

Lee, Y.W.; Chung, M.K.; Kim, S.H.; Park, J.S.; Lee, C.B.; Kim, S.B.; Won, S.Y.; Cho, Y.R.

1983-01-01

The primary objectives of this project are: (1) To review the published information on LOCA/ECCS study (2) To investigate reflood phenomena and to provide necessary information for analytical model development (3) To modyfy and develop a reflood analysis code. To review the published information on LOCA/ECCS, heat transfer phenomena are divided into 4 regions. Heat transfer correlations published in the references are reviewed and classified according to the regions. To investigate reflood phenomena and to provide better modeling of reflood phenomena, experments have been carried out with an electrically heated 3x3 rod bundle. Heat flux and heat transfer coefficients at the hot surface have been determined from the experimental data by HTC program. The influences of the parameters such as flooding rate, coolant subcooling and power generation on the propagation of rewetting front were also investigated. Calculations obtained from REFLUX code were compared with the experimental data to help an understanding of the reflood heat transfer mechanisms, and then some modifications of the code were provided. Improvements in heat transfer correlations of transition and inverted annular film boiling region, and the logic for the selection of heat transfer regime allowed better estimate for rod temperature behavior. (Author)

Numerical studies of the heat-up-phase of Super-Sara 'severe fuel damage'. Boildown tests

International Nuclear Information System (INIS)

Eifler, W.; Shepherd, I.M.

1983-01-01

Calculations to investigate the heat-up phase of the Super-Sara 'severe fuel damage' test matrix have been performed using a simple computer code which models a typical pin. In particular the effect of the exothermic zirconium water reaction on the transient is considered. It is shown that it is possible to achieve the desired objectives of all the tests by a test procedure involving a constant power level a simple flow history. This flow history consists of an initial inlet flow, that has the water saturated at outlet. It is then linearly decreased in a time of the order of 200 seconds to a steady lower value. The clad temperature ramp rate is defined by the power and the peak clad temperature by the ratio of the power of the final steady inlet flow rate. If the final inlet flow rate for a particular power is below a certain critical value then the clad will reach melting temperature. The sensitivity of the results are discussed and a sample calculation is made for each test in the matrix

Probabilistic model fitting for spatio-temporal variability studies of precipitation: the Sara-Brut system - a case study

International Nuclear Information System (INIS)

Dorado Delgado, Jennifer; Burbano Criollo, Juan Carlos; Molina Tabares, Jose Manuel; Carvajal Escobar, Yesid; Aristizabal, Hector Fabio

2006-01-01

In this study, space and time variability of monthly and annual rainfall was analyzed for the downstream influence zone of a Colombian supply-regulation reservoir, Sara-Brut, located on the Cauca valley department. Monthly precipitation data from 18 gauge stations and for a 29-year record (1975-2003) were used. These data were processed by means of time series completion, consistency analyses and sample statistics computations. Theoretical probabilistic distribution models such as Gumbel, normal, lognormal and wake by, and other empirical distributions such as Weibull and Landwehr were applied in order to fit the historical precipitation data set. The fit standard error (FSE) was used to test the goodness of fit of the theoretical distribution models and to choose the best of this probabilistic function. The wake by approach showed the best goodness of fit in 89% of the total gauges taken into account. Time variability was analyzed by means of wake by estimated values of monthly and annual precipitation associated with return periods of 1,052, 1,25, 2, 10, 20 and 50 years. Precipitation space variability is presents by means of ArcGis v8.3 and using krigging as interpolation method. In general terms the results obtained from this study show significant distribution variability in precipitation over the whole area, and particularity, the formation of dry and humid nucleus over the northeastern strip and microclimates at the southwestern and central zone of the study area were observed, depending on the season of year. The mentioned distribution pattern is likely caused by the influence of pacific wind streams, which come from the Andean western mountain range. It is expected that the results from this work be helpful for future planning and hydrologic project design

Identification of new proton-rich rare earth nuclei by means of the coupled system helium jet-isotope separator of SARA

International Nuclear Information System (INIS)

Ollivier, T.

1986-01-01

In order to study new exotic nuclei far from stability we built a fast separation system by coupling a helium jet with the medium-current source of the mass separator. First the tests were made in Lyon and then the system used on line with the heavy ion accelerator SARA, in Grenoble. We obtained efficiency greater than 1% for each element and a better chemical independence. This allowed us to perform experiments on rare-earth region near N=82, with fusion-evaporation reactions after an investigation of various ranges of beam energies. The first results allow to identify two new isotopes, 143 Tb (12s) and 138 Eu (12s). The decay schemes obtained are analysed in the frame of existing models [fr

Land-use evaluation for sustainable construction in a protected area: A case of Sara mountain national park.

Science.gov (United States)

Ristić, Vladica; Maksin, Marija; Nenković-Riznić, Marina; Basarić, Jelena

2018-01-15

The process of making decisions on sustainable development and construction begins in spatial and urban planning when defining the suitability of using land for sustainable construction in a protected area (PA) and its immediate and regional surroundings. The aim of this research is to propose and assess a model for evaluating land-use suitability for sustainable construction in a PA and its surroundings. The methodological approach of Multi-Criteria Decision Analysis was used in the formation of this model and adapted for the research; it was combined with the adapted Analytical hierarchy process and the Delphi process, and supported by a geographical information system (GIS) within the framework of ESRI ArcGIS software - Spatial analyst. The model is applied to the case study of Sara mountain National Park in Kosovo. The result of the model is a "map of integrated assessment of land-use suitability for sustainable construction in a PA for the natural factor". Copyright © 2017 Elsevier Ltd. All rights reserved.

Sensitivity study with respect to direction of ADI method during re-flooding in AHWR

Energy Technology Data Exchange (ETDEWEB)

Kumar, M.; Mukhopadhyay, D. [Bhabha Atomic Research Centre, Mumbai (India). Reactor Safety Div.; Ghosh, A.K. [Bhabha Atomic Research Centre, Mumbai (India). Raja Ramanna Fellow; Kumar, R. [Indian Institute of Technology, Roorkee (India)

2015-05-15

The Advanced Heavy water Reactor (AHWR) is a natural circulation vertical pressure tube type boiling light water cooled and heavy water moderated reactor. As the AHWR fuel bundle quenching under accident condition is designed primarily with radial jets at several axial locations, bottom re-flooding still remain open as another option. Radial direction injection of emergency core cooling leads to rewetting of AHWR fuel cluster in circumferential direction. A 3D fuel pin model has been developed by using Finite Difference Method (FDM) of transient heat conduction equation. Alternating Direction Implicit technique of Finite Difference Method (FDM) has been used for discretisation of numerical equation in different time step at different direction. Sensitivity numerical study with respect to direction of ADI method has been carried out to optimize the time step during the transient as well as steady state and is found that it is insensitivity with direction of solution. Further, to assess influence of circumferential rewetting vis-a-vis axial rewetting. Both the analyses are carried out with same fluid temperature and heat transfer coefficients as boundary conditions. It has been found from the analyses that for radial jet, the circumferential conduction is significant and overall the fuel temperature falls in the quench plane with the initiation of quenching event. The paper discusses the sensitivity study with respect to direction of ADI solution and comparison of numerical results for circumferential and axial rewetting for single pin.

Study on engineering technologies in the Mizunami Underground Research Laboratory. FY 2014. Development of recovery and mitigation technology on excavation damage (Contract research)

International Nuclear Information System (INIS)

Fukaya, Masaaki; Hata, Koji; Akiyoshi, Kenji; Sato, Shin; Takeda, Nobufumi; Miura, Norihiko; Uyama, Masao; Kanata, Tsutomu; Ueda, Tadashi; Hara, Akira; Torisu, Seda; Ishida, Tomoko; Sato, Toshinori; Mikake, Shinichiro; Aoyagi, Yoshiaki

2016-03-01

The researches on engineering technology in the Mizunami Underground Research Laboratory (MIU) project consist of (1) development of design and construction planning technologies, (2) development of construction technology, (3) development of countermeasure technology, (4) development of technology for security and (5) development of technologies for restoration and/or reduction of the excavation damage. As a part of the second phase of the MIU project, research has been focused on the evaluation of engineering technologies including the initial design based on the data obtained during construction. In this research, examination of the plug applied to the future reflood test was conducted as a part of (5) development of technologies for restoration and/or reduction of the excavation damage relating to the engineering technology in the MIU (2014), specifically focused on (1) plug examination (e.g. functions, structure and material) and the quality control methods and (2) analytical evaluation of rock mass behavior around the plug through the reflood test. As a result, specifications of the plug were determined. These specifications should be able to meet requirements for the safety structure and surrounding rock mass against predicted maximum water pressure, temperature stress and seismic force, and for controlling the groundwater inflow, ensuring the access into the reflood gallery and the penetration performance of measurement cable. Also preliminary knowledge regarding the rock mass behavior around the plug after flooding the reflood gallery by installed plug was obtained. A CD-ROM is attached as an appendix. (J.P.N.)

Evaluation report on CCTF core-II reflood test C2 - 8 (Run 67)

International Nuclear Information System (INIS)

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

1987-01-01

In order to study the system pressure effect of the core cooling and flow behavior during the reflood phase of a PWR LOCA, a test was performed with CCTF under the system pressure pf 0.15 MPa as a counterpart test of the CCTF test C2-1 (system pressure 0.42 MPa) and the CCTF test C2-4 (system pressure 0.20 MPa). Through the comparisons of results from these three tests, the following conclusions were obtained: (1) The higher system pressure resulted in the lower temperature rise, the shorter turnaround time and the shorter quench time as observed in the CCTF Core-I system pressure effect tests. (2) The higher system pressure resulted in higher core water head, higher upper plenum water head, higher mass flow rate through the primary loops. On the other hand, the higher system pressure resulted in lower downcomer water head and lower pressure drop through the primary loops and the broken cold leg. These system pressure effects on the flow behavior in the primary system are almost the same as observed in the system pressure effect tests in the CCTF Core-I test series. (3) Before the mixture level in the upper plenum reached the level of the hot leg nozzle, the loop flow resistance coefficient of the intact loops was nearly constant regardless of the system pressure. After the mixture level reached the level of the hot leg nozzle, the loop flow resistance coefficient was increased due to the water accumulation in the hot leg piping and the inlet plenum of the steam generator in these tests. (J.P.N.)

Evaluation report on CCTF core-II reflood test C2-6 (Run 64)

International Nuclear Information System (INIS)

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

1985-03-01

In order to evaluate the effect of the radial power profile on the system behavior and the core thermal hydraulic behavior during the reflood phase of a PWR LOCA, a test was performed using the Cylindrical Core Test Facility(CCTF) with the flat radial power profile. The test was conducted with the same total core power as that of the steep radial power test C2-5(Run 63). Through the comparisons of the results from these two tests, the following conclusions were obtained: (1) The radial power profile in the core has weak effect on the thermal hydraulic behavior in the primary system except the core. (2) Almost the same differential pressure was observed at various elevations in the periphery of the core regardless of different radial power profile. The result suggests that the core differential pressure is determined mainly by the total power and the total stored energy rather than by the local power and the local stored energy. (3) The test results support the single channel core model with the average power rod used in the reactor safety analysis codes such as REFLA-1DS, WREM for the evaluation of the overall system behavior. (4) In the steep radial power test, the heat transfer coefficient in the central(high power) region was higher than that in the peripheral(low power) region. The tendency was not explained by the estimation with the heat transfer correlation developed by Murao and Sugimoto assuming that the void fraction was uniform in a horizontal cross section. It is necessary to study more the dependency of core heat transfer on the radial power profile in the wide core. (author)

An analytical study of thermo-hydrodynamic behaviour of the reflood-phase during a LOCA

International Nuclear Information System (INIS)

Murao, Y.

1977-12-01

The objectives of this study are - the check of the quench model proposed by the author and T. Sudoh, - the establishment of the thermo-hydrodynamics downstream from the quench front, and - the stabilization of the numerical calculations. In order to study these therms, the new version of the reflood analysis code 'REFLA-1D' was developed. The quench modes were classified into the following three types: 1) Liquid column type (rewetting by subcooled water), 2) Dryout type (annular flow type, rewetting by saturated water), and 3) Rewetting type (entire surface temperature higher than rewetting temperature). For the thermo-hydrodynamic model downstream from the quench front, the flow pattern was divided into the five regimes: 1) Subcooled film boiling regime, 2) Transition flow regime, 3) Dispersed flow regime, 4) Superheated steam flow regime, and 5) Rewetted regime. To stabilze the numerical calculation and shorten the computing time, the Lagrangian form of the energy equation of gase phase and dispersed flow region was used instead of the Eulerian form. Considerably close agreement between three PWR-FLECHT tests and the calculated results for the critical Weber number Wec=1.0 was obtained for fuel clad surface temperature and quench time except in earlier stage before turnaround, but poor agreement for the heat transfer characteristics in the transition flow region defined between the quench front and the dispersed flow region. The calculation was relatively stable and the computing time is about the same as a real time for a IBM 370-158 computer. (orig.) [de

Valorización del arte en los textos periodísticos de Clarice Lispector y Sara Gallardo (1967-1973

Directory of Open Access Journals (Sweden)

Guillermina Feudal

2017-12-01

Full Text Available En el contexto regional de democratización en el acceso a los bienes culturales, un aspecto nuclear de la producción periodística de Clarice Lispector y Sara Gallardo revela una sintonía común en el comentario y apropiación de recursos estéticos de las vanguardias “de los sesenta”. La apelación al arte como referencia para evaluar la vida cotidiana constituye una decisión comunicacional que funciona, desde el planteo formal y temático, como eje de las intervenciones críticas que ellas operan en el periodismo. El artículo analiza y compara la fundamentación de las elecciones artísticas de las autoras, así como los universos de discusión en los que desarrollan la contienda estética y debaten sobre los alcances de las nuevas modalidades de interacción cultural.

Dispersed flow film boiling: An investigation of the possibility to improve the models implemented in the NRC computer codes for the reflooding phase of the LOCA

International Nuclear Information System (INIS)

Andreani, M.; Yadigaroglu, G.; Paul Scherrer Inst.

1992-08-01

Dispersed Flow Film Boiling is the heat transfer regime that occurs at high void fractions in a heated channel. The way this heat transfer mode is modelled in the NRC computer codes (RELAP5 and TRAC) and the validity of the assumptions and empirical correlations used is discussed. An extensive review of the theoretical and experimental work related with heat transfer to highly dispersed mixtures reveals the basic deficiencies of these models: the investigation refers mostly to the typical conditions of low rate bottom reflooding, since the simulation of this physical situation by the computer codes has often showed poor results. The alternative models that are available in the literature are reviewed, and their merits and limits are highlighted. The modifications that could improve the physics of the models implemented in the codes are identified

Thermal Hydraulic Tests for Reactor Core Safety

Energy Technology Data Exchange (ETDEWEB)

Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

2007-06-15

The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

Research Implementation and Quality Assurance Project Plan: An Evaluation of Hyperspectral Remote Sensing Technologies for the Detection of Fugitive Contamination at Selected Superfund Hazardous Waste Sites

Science.gov (United States)

Slonecker, E. Terrence; Fisher, Gary B.

2009-01-01

This project is a research collaboration between the U.S. Environmental Protection Agency (EPA) Office of Inspector General (OIG) and the U.S. Geological Survey (USGS) Eastern Geographic Science Center (EGSC), for the purpose of evaluating the utility of hyperspectral remote sensing technology for post-closure monitoring of residual contamination at delisted and closed hazardous waste sites as defined under the Comprehensive Environmental Response Compensation and Liability Act [CERCLA (also known as 'Superfund')] of 1980 and the Superfund Amendments and Reauthorization Act (SARA) of 1986.

Revisiting large break LOCA with the CATHARE-3 three-field model

International Nuclear Information System (INIS)

Valette, Michel; Pouvreau, Jerome; Bestion, Dominique; Emonot, Philippe

2009-01-01

Some aspects of large break LOCA analysis (steam binding, oscillatory reflooding, top-down reflooding) are expected to be improved in advanced system codes from more detailed description of flows by adding a third field for droplets. The future system code CATHARE-3 is under development by CEA and supported by EDF, AREVA-NP and IRSN in the frame of the NEPTUNE project and this paper shows some preliminary results obtained in reflooding conditions. A three-field model has been implemented, including vapor, continuous liquid and liquid droplet fields. This model features a set of nine equations of mass, momentum and energy balance. Such a model allows a more detailed description of the droplet transportation from core to steam generator, while countercurrent flow of continuous liquid is allowed. Code assessment against reflooding experiments in an isolated rod bundle mockup is presented, using 1D meshing of the bundle. Comparisons of CATHARE-3 simulations against data series from PERICLES and RBHT full scale experiments show satisfactory results. Quench front motions are well predicted, as well as clad temperatures in most of the tested runs. The BETHSY 6.7C Integral Effect Test simulating the gravity driven Reflooding process in a scaled PWR circuit is then compared to CATHARE-3 simulation. The three-field model is applied in several parts of the circuit : core, upper plenum, hot leg and steam generator, represented by either 1D or 3D modules, while the classic 6-equation model is used in the other parts of the loop. A short analysis of the results is presented. (author)

Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data

International Nuclear Information System (INIS)

Akimoto, Hajime; Iguchi, Tadashi; Murao, Yoshio

1985-01-01

In the reactor safety assessment during reflood phase of a PWR-LOCA, it is assumed implicitly that the core thermal hydraulic behavior is evaluated by the one-dimensional model with an average power rod. In order to assess the applicability of the one-dimensional treatment, integral tests were performed with various core radial power profiles using the Cylindrical Core Test Facility (CCTF) whose core includes about 2,000 heater rods. The CCTF results confirm that the core radial power profile has weak effect on the thermal hydraulic behavior in the primary system except core. It is also confirmed that the core differential pressure in the axial direction is predicted by the one-dimensional core model with an average power rod even in the case with a steep radial power profile in the core. Even though the core heat transfer coefficient is dependent on the core radial power profile, it is found that the error of the peak clad surface temperature calculation is less than 15 K using the one-dimensional model in the CCTF tests. The CCTF results support the one-dimensional treatment assumed in the reactor safety assessment. (author)

UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

International Nuclear Information System (INIS)

1998-01-01

1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: Investigation of steam/water flow phenomena at the upper tie plate and in the upper plenum and

Severe accident management: a summary of the VAHTI and ROIMA projects

International Nuclear Information System (INIS)

Sairanen, R.

1998-01-01

Two severe accident research projects: 'Severe Accident Management' (VAHTI), 1994-96 and 'Reactor Accidents' Phenomena and Simulation (ROIMA) 1997-98. have been conducted at VTT Energy within the RETU research programme. The main objective was to assist the severe accident management programmes of the Finnish nuclear power plants. The projects had several subtopics. These included thermal hydraulic validation of the APROS code, studies of failure mode of the BWR pressure vessel, investigation of core melt progression within a BWR pressure vessel, containment phenomena, development of a computerised severe accident training tool, and aerosol behaviour experiments. The last topic is summarised by another paper in the seminar. The projects have met the objectives set at the project commencement. Calculation tools have been developed and validated suitable for analyses of questions specific for the Finnish plants. Experimental fission product data have been produced that can be used to validate containment aerosol codes. The tools and results have been utilised in plant assessments. One of the main achievements has been the computer code PASULA for analysis of interactions between core melt and pressure vessel. The code has been applied to pressure vessel penetration analysis. The results have shown the importance of the nozzle construction. Modelling possibilities have recently improved by addition of a creep and porous debris models. Cooling of a degraded BWR core has been systematically studied as joint Nordic projects with a set of severe accident codes. Estimates for coolable conditions have been provided. Recriticality due to reflooding of a damaged core has been evaluated. (orig.)

Impact of 2D/3D-project on LOCA-licensing analysis and reactor safety of PWRs

International Nuclear Information System (INIS)

Winkler, F.; Krebs, W.D.

1989-01-01

In the past LOCA-licensing analysis has included large conservatisms to compensate for the lack of detailed two phase flow and full scale experimental data. The 2D/3D-project was established to improve the data base in order to minimize the conservatisms required. The significant results and findings of the full scale Upper Plenum Test Facility (UPTF) and from the electrically heated Slab Core Test Facility (SCTF) were particularly useful for understanding the multidimensional phenomena in the primary system and in the core of a PWR. UPTF results were used to verify the TRAC-PF1 analysis of a PWR with combined ECC-Injection during the reflood phase of a large break-LOCA. Comparison of these results with results from classic licensing calculations quantifies the large safety margin in earlier licensing procedures and in reactor systems. (orig.)

Notas sobre Feral y las cigüeñas, de Fernando Alonso, y la "Historia del califa cigüeña" (Wilhelm Hauff, Sara Cone Bryant

Directory of Open Access Journals (Sweden)

Hans Christian Hagedorn

2011-01-01

Full Text Available En el presente estudio se analizan las fuentes de la versión de la "Historia del califa cigüeña", incluida en la narración Feral y las cigüeñas (1971, de Fernando Alonso. Para ello se tienen en cuenta el cuento original del autor postromántico alemán Wilhelm Hauff ("Die Geschichte von Kalif Storch", 1825, y la adaptación de este cuento que Sara Cone Bryant realizó para su libro How to tell stories to children (1905, traducción española: El arte de contar cuentos, 1965.

Revisiting large break LOCA with the CATHARE-3 three-field model

International Nuclear Information System (INIS)

Valette, Michel; Pouvreau, Jérôme; Bestion, Dominique; Emonot, Philippe

2011-01-01

Highlights: ► CATHARE 3 enables a three-field analysis of a LB LOCA. ► Reflooding experiments in isolated rod bundles are satisfactory predicted. ► A BETHSY integral test simulation supports the CATHARE 3 3-field assessment. - Abstract: Some aspects of large break LOCA analysis (steam binding, oscillatory reflooding, top-down reflooding) are expected to be improved in advanced system codes from more detailed description of flows by adding a third field for droplets. The future system code CATHARE-3 is under development by CEA and supported by EDF, AREVA-NP and IRSN in the frame of the NEPTUNE project and this paper shows some preliminary results obtained in reflooding conditions. A three-field model has been implemented, including vapor, continuous liquid and liquid droplet fields. This model features a set of nine equations of mass, momentum and energy balance. Such a model allows a more detailed description of the droplet transportation from core to steam generator, while countercurrent flow of continuous liquid is allowed. Code assessment against reflooding experiments in a rod bundle is presented, using 1D meshing of the bundle. Comparisons of CATHARE-3 simulations against data series from PERICLES and RBHT full scale experiments show satisfactory results. Quench front motions are well predicted, as well as clad temperatures in most of the tested runs. The BETHSY 6.7C Integral Effect Test simulating the gravity driven reflooding process in a scaled PWR circuit is then compared to CATHARE-3 simulation. The three-field model is applied in several parts of the circuit: core, upper plenum, hot leg and steam generator, represented by either 1D or 3D modules, while the classic six-equation model is used in the other parts of the loop. An analysis of these first results is presented and future work is defined for improving the droplet behavior simulation in both the upper plenum and the hot legs.

PKL-tests, test series IIB (end of blowdown). Vol. 2

International Nuclear Information System (INIS)

Umminger, K.; Mandl, R.; Nopper, H.; Siemens AG Unternehmensbereich KWU, Erlangen

1987-01-01

As part of the federally subsidized research project 1500 287/A0, the system behavior of a 1300 MWe pressurized water reactor (PWR) was investigated during the depressurization phase (end-of-blowdown, EOB), as well as during the refill and reflood phases of a loss of coolant accident involving a large break in the reactor coolant loop. Appropriate modifications to the system and supplementary instrumentation have made it possible to simulate the EOB (as of 26 bar), the refill phase and reflood phase in sequence. This report includes a detailed description of the instrumentation and the data acquisition system used in Test Series PKL IIB. (orig.) With 6 refs., 2 tabs., 60 figs [de

Instability in newly-established wetlands? Trajectories of floristic change in the re-flooded Hula peatland, northern Israel

Directory of Open Access Journals (Sweden)

D. Kaplan

2012-01-01

Full Text Available Drainage of the 6,000 ha Hula Lake and peatland in northern Israel in the late 1950s caused the loss of a very diverse and rare ecosystem and an important phytogeographic meeting zone for Holarctic and Palaeotropical species. Draining the Hula peatland was only partially successful in creating a large fertile area for cultivation, and in 1994 this led the authorities to re-flood 100 ha of the valley—the Agamon (Agmon—with the aim of rehabilitating the diverse wetland landscape, promoting ecotourism and creating a clear-water body that would contribute to the purification of Lake Kinneret. The vegetation of the restored wetland was monitored for ten years (1997–2006, recording the establishment and abundance of vascular plant species. More than 20 emergent, submerged and riparian species became established. Like a number of other shallow-water wetlands, the Agmon is characterised by considerable ecological fluctuations. This has been expressed in prominent floristic changes in the Agamon since it was created. An increased abundance of Ceratophyllum demersum and Najas minor and a decline in Potamogeton spp., Najas delilei and filamentous algae have been observed. A long-term decline in water level and sediment accumulation has brought about a significant rise in the incidence of Phragmites australis, Typha domingensis and Ludwigia stolonifera in the south-eastern area. A GIS analysis of changes in species dominance shows fluctuations over the years, with only a partial trend of succession towards a P. australis, T. domingensis and L. stolonifera community.

Coloss project

International Nuclear Information System (INIS)

2005-01-01

The COLOSS project was a shared-cost action, co-ordinated by IRSN within the Euratom Research Framework Programme 1998-2002. Started in February 2000, the project lasted three years. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H2 production, melt generation and the source term were studied, through a large number of experiments such as a) dissolution of fresh and high burn-up UO 2 and MOX by molten Zircaloy, b) simultaneous dissolution of UO 2 and ZrO 2 by molten Zircaloy, c) oxidation of U-O-Zr mixtures by steam, d) degradation-oxidation of B 4 C control rods. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics. Break-through were achieved on some issues. Nevertheless, more data are needed for consolidation of the modelling on burn-up effects on UO 2 and MOX dissolution and on oxidation of U-O-Zr and B 4 C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H2 production observed during the reflooding of degraded cores under severe accident conditions. Based on the experimental results obtained on the COLOSS topics, corresponding models were developed and were successfully implemented in several severe accident codes. Upgraded codes were then used for plant calculations to evaluate the consequences of new models on key severe accident sequences occurring in different plants designs involving B 4 C control rods (EPR, BWR, VVER- 1000) as well as in the TMI-2 accident. The large series of plant calculations involved sensitivity studies and code benchmarks. Main severe accident codes in use in the EU for safety studies were used such as ICARE/CATHARE, SCDAP/RELAP5, ASTEC, MELCOR and MAAP4. This activity enabled: a) the assessment of codes to calculate core degradation, b) the identification of main

FLOOD 3 code conversion from Apollo to HP

Energy Technology Data Exchange (ETDEWEB)

Lee, Hae Cho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-01-01

FLOOD3 is a Fortran program used to analyze LOCA Reflood and Post Reflood transients for purpose of mechanistically generating containment sizing mass/energy source terms. The reflood time frame starts when safety injection water first enters the bottom of the core, following the initial LOCA blowdown phase, and the ends when liquid level in the core is high enough to quench the core. The post reflood phase starts at the end of reflood phase and lasts until the end of LOCA transient period. Longterm core cooling is initiated following post reflood phase. FLOOD3 code dose not contain separate capability for longterm cooling analysis. This report firstly describes detailed work carried out for installation of FLOOD3 on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also describes in relation to installation of FLOOD3 on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new.

FLOOD 3 code conversion from Apollo to HP

International Nuclear Information System (INIS)

Lee, Hae Cho

1996-01-01

FLOOD3 is a Fortran program used to analyze LOCA Reflood and Post Reflood transients for purpose of mechanistically generating containment sizing mass/energy source terms. The reflood time frame starts when safety injection water first enters the bottom of the core, following the initial LOCA blowdown phase, and the ends when liquid level in the core is high enough to quench the core. The post reflood phase starts at the end of reflood phase and lasts until the end of LOCA transient period. Longterm core cooling is initiated following post reflood phase. FLOOD3 code dose not contain separate capability for longterm cooling analysis. This report firstly describes detailed work carried out for installation of FLOOD3 on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also describes in relation to installation of FLOOD3 on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new

Evaluation report on CCTF Core-I reflood tests Cl-2 (Run 11) and Cl-3 (Run 12)

International Nuclear Information System (INIS)

Akimoto, Hajime; Murao, Yoshio

1983-06-01

In order to clarify the effect of the initial superheat of the downcomer wall on the system and the core cooling behaviors during the reflood phase of a PWR-LOCA, two tests were performed with the Cylindrical Core Test Facility (CCTF). One is the superheated wall test (test Cl-2) with the initial superheat of 79K, as in the actual PWR, and the other is the saturated wall test (test Cl-3) without any initial superheat. Through the comparisons of the test results from these two tests, the following conclusions were obtained. (1) The initial superheat of the downcomer wall resulted in the lower downcomer water head as observed in our separate-effect tests for the downcomer water head. (2) The superheat also caused the core inlet subcooling to be decreased, and led to the lower core water head. (3) The mass flow rate through the intact loop was reduced only by 4% by the initial superheat of the downcomer wall because the core water head was reduced as well as the downcomer water head. Whereas the mass flow rate through the broken loop was increased because of the increased pressure drop through the broken cold leg. (4) The difference of the core inlet mass flow rate was small between the superheated and the saturated wall tests. It can be considered that small difference of the core inlet mass flow rate results from the compensation of the decreased mass flow rate through the intact loops by the increased mass flow rate through the broken loop. (5) The main discrepancies of the core cooling and the carry-over behaviors between two CCTF tests, were consistent with those observed in the parametric tests for the core inlet subcooling of the FLECHT LOW FLOODING TEST series. (author)

SARA South Observations and Analysis of the Solar Type, Totally Eclipsing, Over Contact Binary, PY Aquarii

Science.gov (United States)

Chamberlain, Heather; Samec, Ronald G.; Caton, Daniel Bruce; Van Hamme, Walter

2018-01-01

PY Aqr (GSC 05191-00853) is a solar Type (T ~ 5750K) eclipsing binary. It was observed in July to October, 2017 at Cerro Tololo in remote mode with the 0.6-m SARA South reflector. Two times of minimum light were calculated from our present observations, a primary and a secondary eclipse:HJD Min I = 2457951.7762±0.0006 HJD Min II = 2458019.5295±00.0003. Both weighted as 1.0.In addition, four timings were determined from online data given in IBVS 5600 and five observations at minima were determined from archived All Sky Automated Survey Data:HJD Min I = 2452908.3165, 2452912.33612 HJD Min II = 2452877.5621, 2452913.34465. All weighted as 0.5.ASAS Observations at minima: 2452094.688, 2453478.882, 2453266.576, 2452093.685 and 54729.600. Each weighted as 0.10The following linear and quadratic ephemerides were determined from all available times of minimum light:JD Hel Min I=2452951.7443±0.0008d + 0.402093441±0.000000099 X E {1} JD Hel Min I=2452951.7439±0.0007d + 0.4020912±0.0000007 X E +0.00000000018 ± 0.00000000006 X E2 {2}A BVRI Bessell filtered simultaneous Wilson-Devinney Program (W-D) solution reveals that the system has a mass ratio of ~0.34 and a component temperature difference of only ~40 K. One low luminosity (Tfact ~ 0.94, ~66 degree radius) large cool region of spots was iterated on the primary component in the WD Synthetic Light Curve computations. It appears in the Southern Hemisphere (colatitude 155 degrees). The Roche Lobe fill-out of the binary is ~17%. The inclination is ~86 degrees. An eclipse duration of ~10 minutes was determined for the primary eclipse and the light curve solution. Additional and more detailed information is given in this report.

1980 Annual status report: super-SARA

International Nuclear Information System (INIS)

1981-01-01

The essential tasks of the JRC for the SUPERSARA project during 1980 were therefore twofold: 1. Actuation of an international Task Force with which to: a. discuss in depth the test objectives and the relationships between the SSTP and the world mosaic of activities in the field of LWR fuel behaviour; b. establish a consensus test matrix; c. identify and discuss the major technological problems affecting the feasibility of attaining the consensus test objectives, especially for the boil-down SFD tests. A Task Force with these objectives was necessary in order to provide the elements for a Council decision on phase II. 2. Conservation of the rythm of the main contractor (UKAEA-Harwell) and subcontractors for the timely fabrication of loop components and the timely design of those new aspects of the plant necessary for the boil-down SFD tests

Sara

International Nuclear Information System (INIS)

Aparo, M.; Dionisi, M.; Vicini, C.; Zeppa, P.; Frazzoli, F.V.; Remetti, R.; Portale, C.

1989-01-01

Nuclear Material Accountability, supported by Containment and Surveillance measures, is a foundamental means for an effective International Safeguard implemention in nuclear plants. Accountability is based on the verification that difference between a material quantity entering a given material balance and the quantity leaving that area in a given period of time, correspond and the amount of material actually present at the moment of the inspection. In the recent years International Safeguards appealing to the needs of timeliness in detecting diversion and concealing activities, devoted ReD efforts on a new Dynamic Accountability procedures (NRTMA) with particular concern with reprocessing plants. The present paper, which is the result of a research activity carried out in the frame of the Italian Support Programme to IAEA for Safeguards implementation, deals with a feasibility study of a NRTMA system to be applied to the EUREX pilot reprocessing plant. Such a feasibility study was performed by developing a computer program based on simulated plant generated data

An experimental investigation of the effect of clad ballooning on the effectiveness of PWR emergency cooling

International Nuclear Information System (INIS)

Cooper, C.A.; Pearson, K.G.; Jowitt, D.

1985-01-01

A series of single phase cooling, forced reflood, gravity reflood and level swell experiments has been performed on a full length, electrically heated 7x7 rod fuel cluster (in the THETIS Rig) containing a blockage simulating very severe clad ballooning. The single phase cooling experiments provided data on the level of heat transfer within the cluster and the enhancement produced by turbulence created by the spacer grids. The forced reflood experiments have led to a better understanding of the rewetting processes during bottom reflooding, the very important influence of spacer grids in two-phase flow and the complicated heat transfer processes within the blockage. The insight obtained has been used to develop a mechanistic model. The gravity reflood experiments investigated steam binding and inlet flow oscillation effects. The variation in inlet flow produced by the variations in system parameters was the dominant influence. The inlet flow oscillations which occurred in gravity reflood appeared not to influence overall rewetting or heat removal performance. The level swell experiments investigated the relationship between void fraction and superficial steam velocity at pressures up to 40 bar and compared the data with various correlations. The correlation of Gardner was found to be most satisfactory. 107 refs., 252 figs.

Heat exchanges during the re-flooding of a water reactor core - within the framework of the 'reference accident'; Echanges thermiques lors du renoyage d'un coeur de reacteur a eau - dans le cadre de 'l'accident de reference'

Energy Technology Data Exchange (ETDEWEB)

Andreoni, Daniel

1975-11-28

After a brief presentation of reported studies made in different countries and regarding the so-called 'reference accident', this research thesis reports the study of reactor re-flooding when the reactor is completely dried and heating elements have reached a temperature between 300 and 900 C, with a constant water flow rate entering the test section, with a constant dissipated electrical power, and by using very simple geometries. After a first part addressing the experimental study, the author reports the development of conduction calculation codes used to compute the flow extracted by the two-phase flow, present the thermal-hydraulic code used to compute local values and to study the correlation of the upstream area exchange coefficient. The author finally reports an analysis of the different existing models and the study of a re-flooding model [French] La presente etude est consacree a l'un des aspects de la surete des reacteurs a eau sous pression, et plus precisement a l'accident tres important qui consiste en une perte de fluide caloporteur (Loss of Coolant Accident - 'LOCA'). Le but de l'etude est de fournir des renseignements necessaires a l'interpretation des experiences effectuees sur des grappes, de donner une correlation de coefficient d'echange dans la zone aval, et de donner aussi un modele de progression du front de trempe pour les analyses de surete. Une etude bibliographique preliminaire nous a permis de faire le point sur les experiences entreprises concernant le refroidissement de secours. Ensuite, les chapitres suivants seront decrits: 1) Le chapitre II, consacre a l'etude experimentale (boucle, sections d'essais, resultats globaux). 2) Le chapitre III ou seront presentes les codes de calcul de conduction, necessaires au calcul du flux extrait par le melange diphasique, le code de thermohydraulique necessaire au calcul des grandeurs locales et l'etude de la correlation du coefficient d'echange de la zone aval. 3) Enfin le chapitre IV ou, apres

Application of thermal hydraulic and severe accident code SOCRAT/V3 to bottom water reflood experiment QUENCH-LOCA-0

International Nuclear Information System (INIS)

Vasiliev, A.D.; Stuckert, J.

2013-01-01

Highlights: ► QLOCA-0 test simulates a design basis LOCA NPP accident with maximum temperature 1300 K. ► Deep understanding of hydraulics and thermal mechanics under accident conditions is necessary. ► We model the test QLOCA-0 with bottom flooding using the Russian code SOCRAT/V3. ► Calculated and experimental data are in a good agreement. ► Experimental procedure is determined to reach a representative LOCA scenario in future tests. -- Abstract: The thermal hydraulic and SFD (severe fuel damage) best estimate computer modeling code SOCRAT/V3 has been used for the calculation of QUENCH-LOCA-0 experiment. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario of the LOCA (loss of coolant accident) nuclear power plant accident sequence in which the overheated up to 1300 K reactor core would be reflooded from the bottom by ECCS (emergency core cooling system). The first test QUENCH-LOCA-0 was successfully conducted at the KIT, Karlsruhe, Germany, in July 22, 2010, and was performed as the commissioning test for this series. The rod claddings are identical to that used in PWRs. The bundle was electrically heated in steam from 800 K to 1340 K with the heat-up rate of approximately 2.7 K/s. After cooling in the saturated steam the bottom flooding with water flow rate of about 100 g/s was initiated. The SOCRAT calculated results are in a good agreement with experimental data taking into account additional quenching due to water condensate entrainment at the steam cooling stage. SOCRAT/V3 has been used for estimation of further steps in experimental procedure to reach a representative LOCA scenario in future tests

SCRELA, LOCA Analysis of Super-Critical Light-Water Reactors

International Nuclear Information System (INIS)

Lee, J.H.; Koshizuka, S.; Oka, Y.

2001-01-01

Description of program or function: LOCA Analysis Code for the Supercritical-Water Cooled Reactor. - Blowdown Module: Calculation of the Blowdown Phase and Refill Phase. - Reflood Module: Calculation of the Reflood Phase

Final report on 3-D experiment project air-water upper plenum experiments

International Nuclear Information System (INIS)

Jacoby, J.K.; Mohr, C.M.

1978-11-01

The results are presented from upper plenum air-water reflood behavior testing performed as part of the program to investigate three-dimensional aspects of PWR LOCA research. Tests described were performed at near ambient temperature and pressure in a plexiglass vessel which included the important features of the upper core and upper plenum regions corresponding to a single fuel bundle in both Westinghouse Electric Corporation (Trojan) and Kraftwerk Union (KKU) PWR designs. The data included observed two-phase flow characteristics, particularly with regard to countercurrent flow, and cinematography of the characteristic upper plenum flow patterns

Project assembling and commissioning of a rewetting test facility

International Nuclear Information System (INIS)

Rezende, H.C.

1985-08-01

A test facility (ITR - Instalacao de Testes de Remolhamento) has been erected at the Thermal-hydraulics Laboratory of CDTN, dedicated to the investigation of the basic phenomena that can occur during the reflood phase of a Loss of Coolant Accident (LOCA) in a Pressurized Water Reactor (PWR), utilizing tubular and annular test sections. The present work consists in a presentation of the facility design and a report of its commissioning. The mechanical aspects of the facility, its power supply system and its instrumentation are described. The results of the instruments calibration and two operational tests are presented and a comparison is done with calculations perfomed usign a computer code. (Author) [pt

Supplmental testimony of the AEC Regulatory Staff. Public rulemaking hearing on: interim acceptance criteria for emergency core cooling systems for light-water cooled power reactors

International Nuclear Information System (INIS)

1972-01-01

Information is presented concerning sensitivity analysis, loop codes, two-phase pressure drop, critical flow model, pump modeling, PWR core flow distribution, accumulator bypass, fuel densification, gap thermal conductance and UO 2 thermal conductivity, transition boiling heat transfer, clad-to-fluid heat transfer, heat transfer at low pressure, reflood rate analyses, containment back pressure during reflood, BWR FLECHT, PWR reflooding heat transfer FLECHT data, embrittlement and post-blowdown loads, fuel rod physico-chemical reactions, flow blockage, small break analysis, and decay heat. (U.S.)

A study of 2-Dimensional effects in the core of a PWR during the refloading phase of a LOCA. Analysis of data of PERICLES experiments with the COBRA-NC code

International Nuclear Information System (INIS)

Reinhardt, H.J.

1989-09-01

The project is embedded in the Shared Cost Action Programme (SCA) of the European Communities (CEC) on Reactor Safety, Research Area No. 4, concerning the analysis of experimental data on loss-of-coolant accidents and emergency core cooling. The PERICLES experiments, performed at CEA in Grenoble, had the objective to study multidimensional effects under well defined conditions concentrating on the inter-assembly character of reflood phenomena. The general aim of the present project is to analyse PERICLES experimental data in order to improve models in relevant system codes. Particular objectives of the project are - the critical evaluation of the experimental data of PERICLES Run 8; - the drawing of conclusions from the data with respect to physical and geometrical models for the multi-bundle reflood analysis; - the performance of one-and multi-dimensional computations with COBRA-NC; - the comparison of computational and experimental data; and - the development of conclusions and specifications of additional research needed. The analysis of the experimetal data of Run 8 was performed by a computer programme developed for postprocessing data of any PERICLES experiment. The postprocessor includes an automatic location of the quenchfront and displays it graphically with respect to time, vertical and horizontal directions. Furthermore, rod and fluid temperatures versus height, quenchtimes versus height, densities versus height, and temperatures, pressures, densities etc. versus time can be plotted. As far as computer simulations are concerned, it was one of the objectives of the present study to analyse in greater detail the multidimensional phenomena during the reflooding phase of a LOCA and to compare the numerical results with the experimental data. Such simulation may serve to adjust and improve existing computer codes as well as to validate the codes. Moreover, computer simulations are able to give information which are not available from experimental data; in the

REPRESENTASI ANTI DISKRIMINASI PADA FILM KARTUN 3D ZOOTOPIA (KAJIAN SEMIOTIKA ROLAND BARTHES

Directory of Open Access Journals (Sweden)

Ali Muqoddas

2016-12-01

Full Text Available Isu diskriminasi SARA seperti tiada habisnya terjadi di dunia ini, begitu juga di Indonesia. Hal ini mengakibatkan tema-tema tentang SARA menjadi sensitif ketika dibahas atau pun difilmkan. Hal ini mengakibatkan jarang ditemui film yang mengangkat tentang SARA. Lain halnya dengan Walt Disney. Walt Disney baru-baru ini merilis film kartun 3D yang mengangkat tema anti diskriminasi SARA yang berjudul Zootopia. Film ini menjadi menarik karena isu diskriminasi SARA yang diangkat dibalut dengan konsep yang kreatif hingga sensivitas isu SARA tersebut menjadi berkurang. Representasi anti diskriminasi SARA pada film Zootopia ini selanjutnya dikaji dengan metode semiotika Roland Barthes dengan pendekatan deskriptif kualitatif. Berdasarkan dari analisis, dapat disimpulkan bahwa film Zootopia memuat pesan ideologi tentang anti diskriminasi SARA bahwa kedudukan manusia dimata manusia yang lain pada hakikatnya adalah sama. Setiap manusia berhak dan wajib memperlakukan dan diperlakukan secara bijak tanpa memperdulikan background asal manusia itu sendiri. Penghargaan pada setiap individu tidak didasarkan pada faktor keturunan, ras, suku ataupun agama, namun didasarkan pada prestasi dari individu itu sendiri. Kata Kunci: diskriminasi SARA, Zootopia, Roland Barthes Abstract The issue of racial discrimination as an endless happen in this world, so also in Indonesia. This resulted in the themes of SARA be sensitive when discussed or filmed. This resulted in a rare film that raised about SARA. Another case with Walt Disney. Walt Disney recently released 3D animated film that mengangat theme of anti-discrimination SARA entitled Zootopia. This film is interesting because racial discrimination issues raised wrapped with a creative concept to the sensitivity of the racial issues be reduced. SARA anti-discrimination representation on film Zootopia is further studied with Roland Barthes semiotic methods with qualitative descriptive approach. Based on the analysis, it can

Coloss project; Le projet Coloss

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

The COLOSS project was a shared-cost action, co-ordinated by IRSN within the Euratom Research Framework Programme 1998-2002. Started in February 2000, the project lasted three years. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H2 production, melt generation and the source term were studied, through a large number of experiments such as a) dissolution of fresh and high burn-up UO{sub 2} and MOX by molten Zircaloy, b) simultaneous dissolution of UO{sub 2} and ZrO{sub 2} by molten Zircaloy, c) oxidation of U-O-Zr mixtures by steam, d) degradation-oxidation of B{sub 4}C control rods. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics. Break-through were achieved on some issues. Nevertheless, more data are needed for consolidation of the modelling on burn-up effects on UO{sub 2} and MOX dissolution and on oxidation of U-O-Zr and B{sub 4}C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H2 production observed during the reflooding of degraded cores under severe accident conditions. Based on the experimental results obtained on the COLOSS topics, corresponding models were developed and were successfully implemented in several severe accident codes. Upgraded codes were then used for plant calculations to evaluate the consequences of new models on key severe accident sequences occurring in different plants designs involving B{sub 4}C control rods (EPR, BWR, VVER- 1000) as well as in the TMI-2 accident. The large series of plant calculations involved sensitivity studies and code benchmarks. Main severe accident codes in use in the EU for safety studies were used such as ICARE/CATHARE, SCDAP/RELAP5, ASTEC, MELCOR and MAAP4. This activity enabled: a) the assessment of codes to calculate core degradation, b) the

Parametric study of recriticality in a boiling water reactor severe accident

International Nuclear Information System (INIS)

Shamoun, B.I.; Witt, R.J.

1994-01-01

Recriticality is possible in a severe accident if unborated or low boron concentration water is added to a damaged core after control rod melting but before fuel melting. Recriticality in a severe accident in a boiling water reactor was parametrically investigated using the TWODANT code. Eigenvalue calculations for a unit central fuel cell with reflective boundary conditions were performed by solving the two-dimensional multigroup steady-state Boltzman transport equation using TWODANT. Two sets of calculations were performed in this work. The first set of calculations was carried out under three types of normal operating conditions to provide reference values for the accident calculations: (a) cold rodded condition, (b) cold unrodded condition, and (c) hot full-power condition. The eigenvalues at these conditions were found to be 1.055, 1.208, and 1.098, respectively. The second set of calculations was carried out after the melting of the control element and during the reflood phase, under the following reflood conditions: (a) reflood with unborated water and (b) reflood with borated water. For the reflood case with unborated water, five values of void fractions were considered (100, 60, 40, 20, and 0%). Decreasing void fractions represent greater refill levels during the reflood process. The system pressure was taken to be 7 MPa, while the moderator temperature was set to 560 K. Plotting the eigenvalue compared with the fraction of control materials lost indicates recriticality is only possible if nearly 100% of the control material is lost from the core. Eigenvalue calculations were repeated for short- and long-term recovery conditions of the reflood phase corresponding to maximum moderator density at 4 MPa pressure and 525 K moderator temperature and for 1 MPa pressure and 325 K moderator temperature, respectively. Recriticality was again observed to be a concern only after losing 95% ore more of control materials from the unit cell

Alteration in Fecal Microbiota Associated with Grain-induced Subacute Ruminal Acidosis Challenge in Dairy Cows

DEFF Research Database (Denmark)

Danscher, Anne Mette; Derakshani, Hooman; Li, Shucong

2014-01-01

in the field are often not detected. Thus, other and better markers of SARA are needed. The purpose of this research was to study the feces microbiome during SARA and assess the possibilities of using feces microbial markers as indicators of SARA. Methods: Six lactating, rumen cannulated, Danish Holstein cows...... were used in a blocked design study including two blocks. In the first block, two cows received control diet and two cows received SARA-challenge diet. In the second block, former control cows received SARA diet while two new cows received control diet. Cows received a total mixed ration (TMR; 24......% concentrate) for four weeks before the trial. SARA was induced by gradual substitution of 40% of TMR with grain pellets (50:50 wheat:barley) over 3 days. Full SARA diet was fed for four days. Rumen pH was measured continuously by indwelling probes (eCow). Feces samples were taken at 9 am and 9 pm on last day...

Experimental studies of thermo-hydraulic processes during passive safety systems operation in new WWER NPP projects

International Nuclear Information System (INIS)

Morozov, A.V.; Remizov, O.V.; Kalyakin, D.S.

2014-01-01

The results of experimental study of thermal-hydraulic processes during operation of the passive safety systems of WWER reactors of new generation are given. The interaction processes of counter flows of saturated steam and cold water in vertical steam-line of the auxiliary passive core reflood system from secondary hydraulic accumulator are studied. The peculiarities of undeveloped boiling on single horizontal tube heating by steam and steam-gas mixture, which is character for WWER steam generator condensing mode, are investigated [ru

The THETIS 80% blocked cluster experiment. Part 6

International Nuclear Information System (INIS)

Pearson, K.G.; Cooper, C.A.; Jowitt, D.

1984-09-01

Thermal-hydraulics experiments on a model PWR fuel assembly containing severe partial blockage are reported. Four types of experiment covered single phase, forced reflood, gravity reflood and level swell studies. A summary of the findings of these experiments is given. (U.K.)

ATHENA-2D: A computer code for simulation of hypothetical recriticality accidents in a thermal neutron spectrum

International Nuclear Information System (INIS)

Schwinkendorf, K.N.

1995-01-01

In a damaged light water reactor core (as in the aftermath of a Three-Mile-Island-like core meltdown), water reflood is performed to carry off decay heat. The severely degraded geometry of the fuel debris bed may increase core reactivity with water reflood. Sufficient boron poisoning of the reflood water is therefore very important. One hypothetical accident is the reintroduction of cooling water that is insufficiently borated, resulting in the damaged reactor attaining criticality in this uncontrolled configuration. The goal in simulating this accident is the prediction of the energy release from the resulting transient

Numerical simulation of AP1000 LBLOCA with SCDAP/RELAP 4.0 code

International Nuclear Information System (INIS)

Xie Heng

2017-01-01

The risk of large-break loss of coolant accident (LBLOCA) is that core will be exposed once the accident occurs, and may cause core damages. New phenomena may occur in LBLOCA due to passive safety injection adopted by AP1000. This paper used SCDAP/RELAP5 4.0 to build the numerical model of AP1000 and double-end guillotine of cold leg is simulated. Reactor coolant system and passive core cooling system were modeled by RELAP5 modular. HEAT STRUCTURE component of RELAP5 was used to simulate the fuel rod. The reflood option in RELAP5 was chosen to be activated or not to study the effect of axial heat conduction. Results show that the axial heat conduction plays an important role in the reflooding phase and can effectively shorten reflood process. An alternative core model is built by SCDAP modular. It is found that the SCDAP model predicts higher maximum peak cladding temperature and longer reflood process than RELAP5 model. Analysis shows that clad oxidation heat plays a key role in the reflood. From the simulation results, it can be concluded that the cladding will keep intact and fission product will not be released from fuel to coolant in LBLOCA. (author)

Investigation of SAM measures during selected MBLOCA sequences along with Station Blackout in a generic Konvoi PWR using ASTECV2.0

International Nuclear Information System (INIS)

Gómez-García-Toraño, Ignacio; Sánchez Espinoza, Víctor Hugo; Stieglitz, Robert

2017-01-01

Highlights: • Reflooding is investigated for selected MBLOCA sequences in a Konvoi PWR using ASTEC. • After SBO, there is a grace time of 40 min up to the detection of a CET = 650 °C. • Major core damage prevented if reflood is launched at CET = 650 °C with 25-40 kg/s. • Values depend on the time when the plant is struck by Station Blackout. • Vessel failure cannot be prevented if supplied mass flow rates are lower than 10 kg/s. - Abstract: The Fukushima accidents have shown that further improvement of Severe Accident Management Guidelines (SAMGs) is necessary for the current fleet of Light Water Reactors. The elaboration of SAMGs requires a broad database of deterministic analyses performed with state-of-the art simulation tools. Within this work, the ASTECV2.0 integral severe accident code is used to study the efficiency of core reflooding (as a SAM measure) during postulated Medium Break LOCA (MBLOCA) scenarios in a German Konvoi PWR. In a first step, the progression of selected MBLOCA sequences without SAM measures has been analysed. The sequences postulate a break in the cold leg of the pressurizer loop and the total loss of AC power at a given stage of the accident. Results show the existence of a 40 min grace time up to the detection of a Core Exit Temperature (CET) of 650 °C providing that the AC power has been maintained at least 1 h after SCRAM. In a second step, an extensive analysis on core reflooding has been carried out. The sequences assume that the plant remains in Station Blackout (SBO) and that reflooding occurs at different times with different mobile pumps. The simulations yield the following results: • Reflooding mass flow rates above 60 kg/s have to be supplied as soon as the CET exceeds 650 °C in order to prevent core melting. • Reflooding mass flow rates ranging from 25–40 kg/s at CET = 650 °C mitigate the accident without major core damage depending on when the plant enters in SBO. • Reflooding mass flow rates lower

PKL/K9, Refill and Reflood Experiment in a Simulated PWR Primary System (PKL)

International Nuclear Information System (INIS)

1981-01-01

reflood process within a pressure vessel considering the influence of the simulated primary loops of a pressurized water reactor (PWR). As to the number of rods and the cross-sections of vessel and components the PKL test-facility is scaled down to 1:134 of a typical West Germany PWR. The elevations and locations have been designed in order to reproduce substantially the original extension. Compared to other cold leg injection tests in the same facility (K5. 3a, K5. 4a) test K9 differs in a uniform radial power profile in the core, an increased temperature of injected Emergency Core Cooling (ECC) water (feed water, from 35 to 53 deg. C), an injection mass flow rate highly reduced to 1:3, and a history of it corresponding to that of a typical US-PWR, on the average somewhat higher maximum initial temperatures in the bundle, and a smaller bundle heating power during the initial phase

Epimural indicator phylotypes of transiently-induced subacute ruminal acidosis in dairy cattle

Directory of Open Access Journals (Sweden)

Stefanie Urimare Wetzels

2016-03-01

Full Text Available The impact of a long-term subacute rumen acidosis (SARA on the bovine epimural bacterial microbiome (BEBM and its consequences for rumen health is poorly understood. This study aimed to investigate shifts in the BEBM during a long-term transient SARA model consisting of two concentrate-diet-induced SARA challenges separated by a one-week challenge break. Eight cows were fed forage and varying concentrate amounts throughout the experiment. In total, 32 rumen papilla biopsies were taken for DNA isolation (4 sampling time points per cow: at the baseline before concentrate was fed, after the first SARA challenge, after the challenge break, and after the second SARA challenge. Ruminal pH was continuously monitored. The microbiome was determined using Illumina MiSeq sequencing of the 16S rRNA gene (V345 region. In total 1,215,618 sequences were obtained and clustered into 6,833 operational taxonomic units (OTUs. Campylobacter and Kingella were the most abundant OTUs (16.5% and 7.1%. According to ruminal pH dynamics, the second challenge was more severe than the first challenge. Species diversity estimates and evenness increased during the challenge break compared to all other sampling time points (P<0.05. During both SARA challenges, Kingella- and Azoarcus-OTUs decreased (0.5 and 0.4 fold-change and a dominant Ruminobacter-OTU increased during the challenge break (18.9 fold-change; P<0.05. qPCR confirmed SARA-related shifts. During the challenge break noticeably more OTUs increased compared to other sampling time points. Our results show that the BEBM re-establishes the baseline conditions slower after a SARA challenge than ruminal pH. Key phylotypes that were reduced during both challenges may help to establish a bacterial fingerprint to facilitate understanding effects of SARA conditions on the BEBM and their consequences for the ruminant host.

Epimural Indicator Phylotypes of Transiently-Induced Subacute Ruminal Acidosis in Dairy Cattle.

Science.gov (United States)

Wetzels, Stefanie U; Mann, Evelyne; Metzler-Zebeli, Barbara U; Pourazad, Poulad; Qumar, Muhammad; Klevenhusen, Fenja; Pinior, Beate; Wagner, Martin; Zebeli, Qendrim; Schmitz-Esser, Stephan

2016-01-01

The impact of a long-term subacute rumen acidosis (SARA) on the bovine epimural bacterial microbiome (BEBM) and its consequences for rumen health is poorly understood. This study aimed to investigate shifts in the BEBM during a long-term transient SARA model consisting of two concentrate-diet-induced SARA challenges separated by a 1-week challenge break. Eight cows were fed forage and varying concentrate amounts throughout the experiment. In total, 32 rumen papilla biopsies were taken for DNA isolation (4 sampling time points per cow: at the baseline before concentrate was fed, after the first SARA challenge, after the challenge break, and after the second SARA challenge). Ruminal pH was continuously monitored. The microbiome was determined using Illumina MiSeq sequencing of the 16S rRNA gene (V345 region). In total 1,215,618 sequences were obtained and clustered into 6833 operational taxonomic units (OTUs). Campylobacter and Kingella were the most abundant OTUs (16.5 and 7.1%). According to ruminal pH dynamics, the second challenge was more severe than the first challenge. Species diversity estimates and evenness increased during the challenge break compared to all other sampling time points (P < 0.05). During both SARA challenges, Kingella- and Azoarcus-OTUs decreased (0.5 and 0.4 fold-change) and a dominant Ruminobacter-OTU increased during the challenge break (18.9 fold-change; P < 0.05). qPCR confirmed SARA-related shifts. During the challenge break noticeably more OTUs increased compared to other sampling time points. Our results show that the BEBM re-establishes the baseline conditions slower after a SARA challenge than ruminal pH. Key phylotypes that were reduced during both challenges may help to establish a bacterial fingerprint to facilitate understanding effects of SARA conditions on the BEBM and their consequences for the ruminant host.

Paducah Gaseous Diffusion Plant site environmental report for 1988

International Nuclear Information System (INIS)

Rogers, J.G.; Jett, T.G.

1989-05-01

Quantities of nonradiological chemical emissions are not included in this report this year. An addendum that will include the information will be published after the Superfund Amendments Reauthorization Act (SARA) Title III report is issued on July 1, 1989. When the addendum is published, probably in late July, a summary of the SARA Title III 313 report will be included. The SARA report provides the community with the opportunity to lean about estimated quantities of certain toxic chemicals used at a facility that are routinely or accidentally released into the environment. The addendum that will be published after the SARA report will summarize the SARA report and is expected to include some additional ''large quantity'' chemicals used or stored at the facilities that are not required to be reported by SARA Title III but are known to be emitted from the facilities. The addendum will not be all inclusive but will provide emissions information on the major chemical emissions to the air, water, or land from processes at the facilities

In vitro residual anti-bacterial activity of difloxacin, sarafloxacin and their photoproducts after photolysis in water

International Nuclear Information System (INIS)

Kusari, Souvik; Prabhakaran, Deivasigamani; Lamshoeft, Marc; Spiteller, Michael

2009-01-01

Fluoroquinolones like difloxacin (DIF) and sarafloxacin (SARA) are adsorbed in soil and enter the aquatic environment wherein they are subjected to photolytic degradation. To evaluate the fate of DIF and SARA, their photolysis was performed in water under stimulated natural sunlight conditions. DIF primarily degrades to SARA. On prolonged photodegradation, seven photoproducts were elucidated by HR-LC-MS/MS, three of which were entirely novel. The residual anti-bacterial activities of DIF, SARA and their photoproducts were studied against a group of pathogenic strains. DIF and SARA revealed potency against both Gram-positive and -negative bacteria. The photoproducts also exhibited varying degrees of efficacies against the tested bacteria. Even without isolating the individual photoproducts, their impact on the aquatic environment could be assessed. Therefore, the present results call for prudence in estimating the fate of these compounds in water and in avoiding emergence of resistance in bacteria caused by the photoproducts of DIF and SARA. - Assessment of the residual anti-bacterial efficacies of difloxacin, sarafloxacin and their photoproducts in water, and estimating their impact on the aquatic environment in inducing resistance to microorganisms.

PWR FLECHT SEASET 21-rod bundle flow blockage task. Task plan report. FLECHT SEASET Program report No. 5

International Nuclear Information System (INIS)

Hochreiter, L.E.; Basel, R.A.; Dennis, R.J.; Lee, N.; Massie, H.W. Jr.; Loftus, M.J.; Rosal, E.R.; Valkovic, M.M.

1980-10-01

This report presents a descriptive plan of tests for the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). This task will consist of forced and gravity reflooding tests utilizing electrical heater rods to simulate PWR nuclear core fuel rod arrays. All tests will be performed with a cosine axial power profile. These tests are planned to be used to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 161-rod flow blockage bundle tests

Evaluation report on CCTF core-I reflood tests Cl-5 (Run 14), Cl-10 (Run 19) and Cl-12 (Run 21)

International Nuclear Information System (INIS)

Akimoto, Hajime; Murao, Yoshio

1983-06-01

Three tests Cl-5 (Run 14), Cl-10 (Run 19) and Cl-12 (Run 21) were performed using the Cylindrical Core Test Facility to study the effect of the containment pressure on the core cooling and the system behaviors during the reflood phase of a PWR-LOCA. The containment pressures of these tests were 0.15, 0.20 and 0.30 MPa for the tests Cl-10, Cl-5 and Cl-12, respectively. Through the comparison of the test results from these three tests, the following results were obtained. (1) The higher containment pressure gave the higher heat transfer coefficient in the core. This resulted in the lower turnaround temperature, the shorter turnaround time and the shorter quench time at the higher containment pressure. (2) In the higher containment pressure test, the higher core water head, the higher upper plenum water head, the higher downcomer water head in the early period and the lower downcomer water head in the later period were observed than those in the lower containment pressure test. This resulted in the higher pressure drop through the intact loop in the early period of the tests and the lower pressure drop in the later period of the test with the containment pressure. (3) The pressure drop through the broken cold leg pressurized the primary system. The pressure drop through the broken cold leg was decreased with the containment pressure. (4) The core inlet mass flow rate was increased with the containment pressure as observed in the FLECHT-SET phase B1 test. In quantity, however, the effect of the containment pressure on the increase of the core inlet mass flow rate was less in the CCTF than that in the FLECHT-SET. The less sensitivity in the CCTF was attributed mainly to the great pressure drop through the broken cold leg, which was not observed in the FLECHT-SET with big broken cold leg. (5) The system effect of the containment pressure was explained quantitatively. (author)

Mapping the Heart

Science.gov (United States)

Hulse, Grace

2012-01-01

In this article, the author describes how her fourth graders made ceramic heart maps. The impetus for this project came from reading "My Map Book" by Sara Fanelli. This book is a collection of quirky, hand-drawn and collaged maps that diagram a child's world. There are maps of her stomach, her day, her family, and her heart, among others. The…

The use of sodic monensin and probiotics for controlling subacute ruminal acidosis in sheep

Directory of Open Access Journals (Sweden)

Elizabeth Schwegler

2015-02-01

Full Text Available The aim of this work was to validate a protocol for induction of subacute ruminal acidosis (SARA (Experiment 1 and test the efficiency of probiotic Saccharomyces cerevisiae or monensin to avoid pH ruminal drops in sheep (Experiment 2. In Experiment 1, six ewes were fasted for two days and then fed most with concentrate during four days. Ewes in this protocol had ruminal fluid pH below 6.0 and kept it for 75 consecutive hours. In Experiment 2, 18 sheep were distributed into three groups: Control (CG, n = 6, monensin (MG, n = 6 and probiotic group (PG, n = 6. SARA was induced according Experiment 1. PG had lower pH (5.7 ± 0.1 than CG (6.0 ± 0.1 (P = 0.05, while MG (5.7 ± 0.1 was similar to both during SARA induction. SARA induction reduced ruminal protozoa population (P < 0.05 and increased chloride concentrations in ruminal fluid (P < 0.01. In serum, SARA increased concentrations of phosphorus (P < 0.01, AST (P < 0.01 and GGT (P < 0.01, but reduced LDH (P < 0.01. In conclusion, the protocol used for SARA induction was able to maintain ruminal pH between 5.5-6.0 for more than 48 hours. However, monensin and probiotics supplementation was not effective in preventing changes in ruminal and serum parameters during SARA.

Heat transfer in rod bundles with severe clad deformations

International Nuclear Information System (INIS)

Ihle, P.

1984-04-01

The content of the paper is focused on heat transfer conditions during the reflood phase of a LOCA in slightly to severely deformed PWR fuel rod bundle geometries. The status of analytical and, especially, of experimental work is described as far as it is possible within this frame. Emphasis is placed on the presentation of the results of ''Flooding Experiments with Blocked Arrays'' (FEBA), a program performed at the Kernforschungszentrum Karlsruhe in the frame of the Project Nuclear Safety (PNS). (orig./WL) [de

The THETIS 80% blocked cluster experiment. Part 1

International Nuclear Information System (INIS)

Jowitt, D.; Cooper, C.A.; Pearson, K.G.

1984-09-01

Thermal-hydraulics experiments on a model PWR fuel assembly containing severe partial blockage are reported. Four types of experiment covered single phase, forced reflood, gravity reflood and level swell studies. The main features of the THETIS rig are described together with its instrumentation. Test cluster and blockage geometry are described in detail. Problems encountered during the tests are outlined. (U.K.)

Posttest analysis of international standard problem 10 using RELAP4/MOD7

International Nuclear Information System (INIS)

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.; Behling, S.R.

1981-01-01

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of a hypothetical loss-of-coolant accident

Thermal-Hydraulic Tests for Reactor Core Safety

International Nuclear Information System (INIS)

Chun, Se Young; Chung, Moon Ki; Baek, Won Pil and others

2005-04-01

The reflood experiments for single rod annulus geometry have been performed to investigate the effect of spacer grid on thermal-hydraulics under reflood conditions. The reflood experimental loop for 6x6 rod bundle with a spacer grid developed in Korea has been provided. About 8000 data points for Post-CHF heat transfer have been obtained from the experiments About 1400 CHF data points for 3x3 Water and 5x5 Freon rod bundles have been obtained. The existing evaluation methodology for core safety under return-to-power conditions has been investigated using KAERI low flow CHF database. The hydraulic tests for turbulence mixing characteristics in subchannel of 5x5 rod bundle have been carried out using advanced measurement technique, LVD and the database for various spacer grids have been provided. In order to measure the turbulence mixing characteristics in details, the hydraulic loop with a magnified 5x5 rod bundle has been prepared. The database which was constructed through a systematic thermal hydraulic tests for the reflood phenomenon, CHF, Post-CHF is surely to be useful to the industry field, the regulation body and the development of thermal-hydraulic analysis code

Recent results from CEC cost sharing research programme on LWR fuel behaviour under accident conditions

International Nuclear Information System (INIS)

Fairbairn, S.A.

1983-01-01

The present structure and intentions of the CEC sponsored cost sharing programme for LWR safety research are outlined. Detailed results are reported for two projects from this programme. The first project concerns experimental data on the thermohydraulic effects of flow diversion around ballooned fuel rods. Data are presented on single and two phase heat transfer in an electrically heated rod bundle. Detailed photographic data on droplet behaviour are also given. The second project is an investigation of the effects of zircaloy oxidation on rewetting during reflood. It is shown that as oxide thickness increases from 1μm to 76μm that rewet rates can increase by up to 40%. A systematic effect of oxidation on rewet temperatures is also noted. (author)

An Investigation into Rumen Fungal and Protozoal Diversity in Three Rumen Fractions, during High-Fiber or Grain-Induced Sub-Acute Ruminal Acidosis Conditions, with or without Active Dry Yeast Supplementation.

Science.gov (United States)

Ishaq, Suzanne L; AlZahal, Ousama; Walker, Nicola; McBride, Brian

2017-01-01

Sub-acute ruminal acidosis (SARA) is a gastrointestinal functional disorder in livestock characterized by low rumen pH, which reduces rumen function, microbial diversity, host performance, and host immune function. Dietary management is used to prevent SARA, often with yeast supplementation as a pH buffer. Almost nothing is known about the effect of SARA or yeast supplementation on ruminal protozoal and fungal diversity, despite their roles in fiber degradation. Dairy cows were switched from a high-fiber to high-grain diet abruptly to induce SARA, with and without active dry yeast (ADY, Saccharomyces cerevisiae ) supplementation, and sampled from the rumen fluid, solids, and epimural fractions to determine microbial diversity using the protozoal 18S rRNA and the fungal ITS1 genes via Illumina MiSeq sequencing. Diet-induced SARA dramatically increased the number and abundance of rare fungal taxa, even in fluid fractions where total reads were very low, and reduced protozoal diversity. SARA selected for more lactic-acid utilizing taxa, and fewer fiber-degrading taxa. ADY treatment increased fungal richness (OTUs) but not diversity (Inverse Simpson, Shannon), but increased protozoal richness and diversity in some fractions. ADY treatment itself significantly ( P PERMANOVA, P = 0.0001, P = 0.0452, P = 0.0068, Monte Carlo correction, respectively, and location was a significant factor ( P = 0.001, Monte Carlo correction) for protozoa. Diet-induced SARA shifts diversity of rumen fungi and protozoa and selects against fiber-degrading species. Supplementation with ADY mitigated this reduction in protozoa, presumptively by triggering microbial diversity shifts (as seen even in the high-fiber diet) that resulted in pH stabilization. ADY did not recover the initial community structure that was seen in pre-SARA conditions.

Problems of two-phase flows in water cooled and moderated reactors

International Nuclear Information System (INIS)

Syu, Yu.

1984-01-01

Heat exchange in two-phase flows of coolant in loss of coolant accidents in PWR and BWR reactors has been investigated. Three main stages of accident history are considered: blowdown, reflooding using emergency core cooling system and rewetting. Factors, determining the rate of coolant leakage and the rate of temperature increase in fuel cladding during blowdown, processes of vapour during reflooding and liquid priming by vapour during rewetting, are discussed

In-Vessel Coolability. Workshop Proceedings, in collaboration with EC-SARNET

International Nuclear Information System (INIS)

2011-01-01

Severe Accident Management Guidelines increase focus on containment integrity after some progression in the course of a severe accident. This change in priorities is made according to criteria that vary depending on reactor type and specific procedures. Once a water source has been recovered, different accident management strategies can be used: send water into the core and/or cool the reactor pressure vessel (RPV) externally. It should be noticed that, depending on the amount of water available, these strategies might conflict with other uses of water such as for instance activating spray systems in the containment or may have deleterious effects as for instance an increase in the production of hydrogen. Generally, for in-vessel reflooding, the models used for evaluation of accident management measures suffer from a lack of validation. Given this background, the objectives of the workshop were: -) to exchange information on different Severe Accident Management strategies used or contemplated for the in-vessel coolability issue; -) to review recent, ongoing and planned experimental programmes on reflooding; -) to review models used for reflooding in severe accident calculation tools, either simplified or sophisticated; -) to exchange information on the treatment of reflooding in different safety studies such as Probabilistic Safety Assessment; and -) to provide recommendations for future work, as necessary

SPACE Code Assessment for FLECHT Test

Energy Technology Data Exchange (ETDEWEB)

Ahn, Hyoung Kyoun; Min, Ji Hong; Park, Chan Eok; Park, Seok Jeong; Kim, Shin Whan [KEPCO E and C, Daejeon (Korea, Republic of)

2015-10-15

According to 10 CFR 50 Appendix K, Emergency Core Cooling System (ECCS) performance evaluation model during LBLOCA should be based on the data of FLECHT test. Heat transfer coefficient (HTC) and Carryout Rate Fraction (CRF) during reflood period of LBLOCA should be conservative. To develop Mass and Energy Release (MER) methodology using Safety and Performance Analysis CodE (SPACE), FLECHT test results were compared to the results calculated by SPACE. FLECHT test facility is modeled to compare the reflood HTC and CRF using SPACE. Sensitivity analysis is performed with various options for HTC correlation. Based on this result, it is concluded that the reflood HTC and CRF calculated with COBRA-TF correlation during LBLOCA meet the requirement of 10 CFR 50 Appendix K. In this study, the analysis results using SPACE predicts heat transfer phenomena of FLECHT test reasonably and conservatively. Reflood HTC for the test number of 0690, 3541 and 4225 are conservative in the reference case. In case of 6948 HTC using COBRATF is conservative to calculate film boiling region. All of analysis results for CRF have sufficient conservatism. Based on these results, it is possible to apply with COBRA-TF correlation to develop MER methodology to analyze LBLOCA using SPACE.

The MERLIN programme: Pt. 7

International Nuclear Information System (INIS)

Worswick, D.; Melvin, G.T.; Walker, J.C.

1989-08-01

This report describes a series of experiments carried out in the MERLIN rig to provide information about the vibration response of a fuel rod during the reflood stage of a large break Loss of Coolant Accident (LOCA) in a PWR. Vibration of fuel rods in a LOCA is expected to influence the stability and radial position of the fuel stack within the cladding, and these factors are known to have a considerable influence on cladding deformation behaviour. The MERLIN rig at SL is intended to investigate the deformation behaviour of Zircaloy fuel rod cladding under conditions approximating those of a large break LOCA. An assembly of electrically heated fuel rod simulators (6X6 cluster) is subjected to a temperature transient simulating that predicted to occur in a LOCA, including the initiation of bottom reflooding at a suitable stage. As part of the preliminary test programme to characterise the rig, vibration measurements were made during a series of experiments in which a flat top transient (steady state condition) was achieved during reflood cooling. It was found that the vibration response was not sensitive to cladding temperature (500 - 650 0 C), but was dependent on the time into reflood for early stages of the transient. (author)

Epimural Indicator Phylotypes of Transiently-Induced Subacute Ruminal Acidosis in Dairy Cattle

OpenAIRE

Wetzels, Stefanie U.; Mann, Evelyne; Metzler-Zebeli, Barbara U.; Pourazad, Poulad; Qumar, Muhammad; Klevenhusen, Fenja; Pinior, Beate; Wagner, Martin; Zebeli, Qendrim; Schmitz-Esser, Stephan

2016-01-01

The impact of a long-term subacute rumen acidosis (SARA) on the bovine epimural bacterial microbiome (BEBM) and its consequences for rumen health is poorly understood. This study aimed to investigate shifts in the BEBM during a long-term transient SARA model consisting of two concentrate-diet-induced SARA challenges separated by a 1-week challenge break. Eight cows were fed forage and varying concentrate amounts throughout the experiment. In total, 32 rumen papilla biopsies were taken for DNA...

Loglines. May-June 2014

Science.gov (United States)

2014-06-01

Kathleen T. Rhem Editor: Jacob Boyer Layout/Design: Paul Henry Crank Writers: Beth Reece Sara Moore Amanda Neumann Loglines is prepared...www.logisticsinformationservice.dla.mil dla.tng@dla.mil Jake Logan , a maintenance employee from the U.S. Army Corps of Engineers’ Greers Ferry Project...of their expertise. “The Joint Petroleum Seminar is designed to gather joint petroleum Defense, both incoming and incumbent, and expose them to

Environmental compliance at U.S. Department of Energy FUSRAP (Formerly Utilized Sites Remedial Action Program) sites

International Nuclear Information System (INIS)

Liedle, S.D.; Clemens, B.W.

1988-01-01

With the promulgation of the Superfund Amendments and Reauthorization Act (SARA), federal facilities were required to comply with the Comprehensive Environmental Response Compensation and Liability Act (CERCLA) in the same manner as any non-government entity. This presented challenges for the Department of Energy (DOE) and other federal agencies involved in remedial action work because there are many requirements under SARA that overlap other laws requiring DOE compliance, e.g., the National Environmental Policy Act (NEPA). This paper outlines the options developed to comply with CERCLA and NEPA as part of active, multi-site remedial action program. The program, the Formerly Utilized Sites Remedial Action Program (FUSRAP), was developed to identify, clean up, or control sites containing residual radioactive or chemical contamination as a result of the nation's early development of nuclear power. During the Manhattan Project, uranium was extracted from ores and resulted in mill concentrates, purified metals, and waste products that were transported for use or disposal at other locations. Figure 1 shows the steps for producing uranium metal during the Manhattan Project. As a result of these activities materials, equipment, buildings, and land became contaminated, primarily with naturally occurring radionuclides. Currently, FUSRAP includes 29 sites; three are on the Environmental Protection Agency's (EPA's) National Priorities List (NPL) of hazardous waste sites

FUSRAP adapts to the amendments of Superfund

International Nuclear Information System (INIS)

Atkin, R.G.; Liedle, S.D.; Clemens, B.W.

1988-01-01

With the promulgation of the Superfund Amendments and Reauthorization Act (SARA) federal facilities were required to comply with the Comprehensive Environmental Response Compensation and Liability Act (CERCLA) in the same manner as any non-government entity. This situation presented challenges for the Department of Energy (DOE) and other federal agencies involved in remedial action work because of the requirements under SARA that overlap other laws requiring DOE compliance, e.g., the National Environmental Policy Act (NEPA). This paper outlines options developed to comply with CERCLA and NEPA as part of an active, multi-site remedial action program. The program, the Formerly Utilized Sites Remedial Action Program (FUSRAP), was developed to identify, clean up, or control sites containing residual radioactive contamination resulting from the nation's early development of nuclear power. During the Manhattan Project, uranium was extracted from domestic and foreign ores and resulted in mill concentrates, purified metals, and waste products that were transported for use or disposal at other locations. Figure 1 shows the steps for producing uranium metal during the Manhattan Project. As a result of these activities materials equipment, buildings, and land became contaminated, primarily with naturally occurring radionuclides. Currently, FUSRAP includes 29 sites; three are on the Environmental Protection Agency's (EPA's) National Priorities List (NPL) of hazardous waste sites

Interpretation of Ersec tests on the backup cooling of pressurized water reactors, by using the FLIRA code

International Nuclear Information System (INIS)

Reviglio, Christiane

1977-01-01

This research thesis addresses the study of the most severe accident, or reference accident, which might occur in nuclear reactors, a clean break of a cold branch of the primary circuit, which may put the integrity of barriers against radioactive products dispersion outside of the reactor into question again. More particularly, the thesis addresses the study of the backup cooling system, and the fact that fluid flow during re-flooding must be predicted, and that heat exchange coefficients must be known in order to assess the evolution of sheath temperatures. The research comprised an experimental part which aimed at reproducing as faithfully as possible the re-flooding sequence on a tube with internal flow or on a cluster for a better core simulation. These are the ERSEC tests which are to be interpreted. It also comprised a theoretical part based on the use of computational codes which simulate the different phases of the accident and of backup fluid injection. These codes are based on physical models which describe two-phase flows and heat exchanges, and are adjusted to experimental results. The FLIRA code is used which simulates the re-flooding of a reactor duct, and determines the evolution of different values (pressure, temperatures, flow rate, and so on) during the re-flooding process. Thus, the author presents the reference accident, reports studies performed in the USA and in France (ERSEC tests), indicates the various flow regimes and describes heat exchange mechanisms during re-flooding, presents ERSEC test results, presents the FLIRA code, reports the elaboration of governing equations, indicates the various models introduced in the FLIRA code, and describes the numerical processing of equations. He finally gives a first interpretation of ERSEC tests based on the use of the FLIRA code

Annual report 1979

International Nuclear Information System (INIS)

1980-01-01

The research and development activities of the ISN for 1979 are reported. The main thrust is directed towards the SARA (Rhone-Alpes Accelerator System) project and the various technical and experimental aspects. Research work was carried out in the following areas: nuclear spectroscopy, giant resonances, reaction processes, experiments at ILL, intermediate energy physics and for the theoretical group: nuclear structure, nuclear reactions, three-body problem, thermal neutron capture and hypernuclear spectroscopy [fr

Screamy Bird

DEFF Research Database (Denmark)

Tarby, Sara; Cermak, Daniel

2016-01-01

Sara Tarby, Daniel Cermak-Sassenrath. Screamy Bird. Digital game. Kulturnatten 2016, Danish Science Ministry, Copenhagen, DK, Oct 14, 2016.......Sara Tarby, Daniel Cermak-Sassenrath. Screamy Bird. Digital game. Kulturnatten 2016, Danish Science Ministry, Copenhagen, DK, Oct 14, 2016....

Testing methodologies for quantifying physical models uncertainties. A comparative exercise using CIRCE and IPREM (FFTBM)

Energy Technology Data Exchange (ETDEWEB)

Freixa, Jordi, E-mail: jordi.freixa-terradas@upc.edu; Alfonso, Elsa de, E-mail: elsa.de.alfonso@upc.edu; Reventós, Francesc, E-mail: francesc.reventos@upc.edu

2016-08-15

Highlights: • Uncertainty of physical models are a key issue in Best estimate plus uncertainty analysis. • Estimation of uncertainties of physical models of thermal hydraulics system codes. • Comparison of CIRCÉ and FFTBM methodologies. • Simulation of reflood experiments in order to evaluate uncertainty of physical models related to the reflood scenario. - Abstract: The increasing importance of Best-Estimate Plus Uncertainty (BEPU) analyses in nuclear safety and licensing processes have lead to several international activities. The latest findings highlighted the uncertainties of physical models as one of the most controversial aspects of BEPU. This type of uncertainties is an important contributor to the total uncertainty of NPP BE calculations. Due to the complexity of estimating this uncertainty, it is often assessed solely by engineering judgment. The present study comprises a comparison of two different state-of-the-art methodologies CIRCÉ and IPREM (FFTBM) capable of quantifying the uncertainty of physical models. Similarities and differences of their results are discussed through the observation of probability distribution functions and envelope calculations. In particular, the analyzed scenario is core reflood. Experimental data from the FEBA and PERICLES test facilities is employed while the thermal hydraulic simulations are carried out with RELAP5/mod3.3. This work is undertaken under the framework of PREMIUM (Post-BEMUSE Reflood Model Input Uncertainty Methods) benchmark.

An Investigation into Rumen Fungal and Protozoal Diversity in Three Rumen Fractions, during High-Fiber or Grain-Induced Sub-Acute Ruminal Acidosis Conditions, with or without Active Dry Yeast Supplementation

Directory of Open Access Journals (Sweden)

Suzanne L. Ishaq

2017-10-01

Full Text Available Sub-acute ruminal acidosis (SARA is a gastrointestinal functional disorder in livestock characterized by low rumen pH, which reduces rumen function, microbial diversity, host performance, and host immune function. Dietary management is used to prevent SARA, often with yeast supplementation as a pH buffer. Almost nothing is known about the effect of SARA or yeast supplementation on ruminal protozoal and fungal diversity, despite their roles in fiber degradation. Dairy cows were switched from a high-fiber to high-grain diet abruptly to induce SARA, with and without active dry yeast (ADY, Saccharomyces cerevisiae supplementation, and sampled from the rumen fluid, solids, and epimural fractions to determine microbial diversity using the protozoal 18S rRNA and the fungal ITS1 genes via Illumina MiSeq sequencing. Diet-induced SARA dramatically increased the number and abundance of rare fungal taxa, even in fluid fractions where total reads were very low, and reduced protozoal diversity. SARA selected for more lactic-acid utilizing taxa, and fewer fiber-degrading taxa. ADY treatment increased fungal richness (OTUs but not diversity (Inverse Simpson, Shannon, but increased protozoal richness and diversity in some fractions. ADY treatment itself significantly (P < 0.05 affected the abundance of numerous fungal genera as seen in the high-fiber diet: Lewia, Neocallimastix, and Phoma were increased, while Alternaria, Candida Orpinomyces, and Piromyces spp. were decreased. Likewise, for protozoa, ADY itself increased Isotricha intestinalis but decreased Entodinium furca spp. Multivariate analyses showed diet type was most significant in driving diversity, followed by yeast treatment, for AMOVA, ANOSIM, and weighted UniFrac. Diet, ADY, and location were all significant factors for fungi (PERMANOVA, P = 0.0001, P = 0.0452, P = 0.0068, Monte Carlo correction, respectively, and location was a significant factor (P = 0.001, Monte Carlo correction for protozoa

En ny flugt ind i skoven

DEFF Research Database (Denmark)

Haarder, Jon Helt

2008-01-01

Thure Erik Lund: IND. (Inn). Oversat fra norsk af Sara Koch. 192 sider, 199 kr. Gyldendal. UDKOMMER I DAG. Fire stjerner......Thure Erik Lund: IND. (Inn). Oversat fra norsk af Sara Koch. 192 sider, 199 kr. Gyldendal. UDKOMMER I DAG. Fire stjerner...

Two-phase flow dynamics in ECC

International Nuclear Information System (INIS)

Albraaten, P.J.

1981-07-01

The present report summarizes the achievements within the project ''Two-phase Systems and ECC''. The results during 1978 - 1980 are accounted for in brief as they have been documented in earlier reports. The results during the first half of 1981 are accounted for in greater detail. They contain a new model for the Basset force and test runs with this model using the test code RISQUE. Furthermore, test runs have been performed with TRAC-PD2 MOD 1. This code was implemented on Edwards Pipe Blowdown experiment (a standard test case) and UC-Berkeley Reflooding experiment (a non-standard test case.) (Auth.)

Butanol/Gasoline Test Plan

Science.gov (United States)

2012-03-01

authorities. METHODS FOR CLEANING UP: Contain spillage and then collect with non-combustible absorbent material, (e.g. sand, soil , diatomaceous earth...SARA 311/312 Hazards: Fire Hazard, Acute Health Hazard, and Chronic Health Hazard CERCLA SECTION 103 and SARA SECTION 304 (RELEASE TO THE ENVIROMENT

A perspective of PC-based probabilistic risk assessment

International Nuclear Information System (INIS)

Sattison, M.B.; Rasmuson, D.M.; Robinson, R.C.; Russell, K.D.; Van Siclen, V.S.

1987-01-01

Probabilistic risk assessment (PRA) information has been under-utilized in the past due to the large effort required to input the PRA data and the large expense of the computers needed to run PRA codes. The microcomputer-based Integrated Reliability and Risk Analysis System (IRRAS) and the System Analysis and Risk Assessment (SARA) System, under development at the Idaho National Engineering Laboratory, have greatly enhanced the ability of managers to use PRA techniques in their decision-making. IRRAS is a tool that allows an analyst to create, modify, update, and reanalyze a plant PRA to keep the risk assessment current with the plant's configuration and operation. The SARA system is used to perform sensitivity studies on the results of a PRA. This type of analysis can be used to evaluate proposed changes to a plant or its operation. The success of these two software projects demonstrate that risk information can be made readily available to those that need it. This is the first step in the development of a true risk management capability

Substorm associated radar auroral surges: a statistical study and possible generation model

Directory of Open Access Journals (Sweden)

B. A. Shand

Full Text Available Substorm-associated radar auroral surges (SARAS are a short lived (15–90 minutes and spatially localised (~5° of latitude perturbation of the plasma convection pattern observed within the auroral E-region. The understanding of such phenomena has important ramifications for the investigation of the larger scale plasma convection and ultimately the coupling of the solar wind, magnetosphere and ionosphere system. A statistical investigation is undertaken of SARAS, observed by the Sweden And Britain Radar Experiment (SABRE, in order to provide a more extensive examination of the local time occurrence and propagation characteristics of the events. The statistical analysis has determined a local time occurrence of observations between 1420 MLT and 2200 MLT with a maximum occurrence centred around 1700 MLT. The propagation velocity of the SARAS feature through the SABRE field of view was found to be predominately L-shell aligned with a velocity centred around 1750 m s^–1 and within the range 500 m s^–1 and 3500 m s^–1. This comprehensive examination of the SARAS provides the opportunity to discuss, qualitatively, a possible generation mechanism for SARAS based on a proposed model for the production of a similar phenomenon referred to as sub-auroral ion drifts (SAIDs. The results of the comparison suggests that SARAS may result from a similar geophysical mechanism to that which produces SAID events, but probably occurs at a different time in the evolution of the event.

Key words. Substorms · Auroral surges · Plasma con-vection · Sub-auroral ion drifts

RNA structure alignment by a unit-vector approach.

Science.gov (United States)

Capriotti, Emidio; Marti-Renom, Marc A

2008-08-15

The recent discovery of tiny RNA molecules such as microRNAs and small interfering RNA are transforming the view of RNA as a simple information transfer molecule. Similar to proteins, the native three-dimensional structure of RNA determines its biological activity. Therefore, classifying the current structural space is paramount for functionally annotating RNA molecules. The increasing numbers of RNA structures deposited in the PDB requires more accurate, automatic and benchmarked methods for RNA structure comparison. In this article, we introduce a new algorithm for RNA structure alignment based on a unit-vector approach. The algorithm has been implemented in the SARA program, which results in RNA structure pairwise alignments and their statistical significance. The SARA program has been implemented to be of general applicability even when no secondary structure can be calculated from the RNA structures. A benchmark against the ARTS program using a set of 1275 non-redundant pairwise structure alignments results in inverted approximately 6% extra alignments with at least 50% structurally superposed nucleotides and base pairs. A first attempt to perform RNA automatic functional annotation based on structure alignments indicates that SARA can correctly assign the deepest SCOR classification to >60% of the query structures. The SARA program is freely available through a World Wide Web server http://sgu.bioinfo.cipf.es/services/SARA/. Supplementary data are available at Bioinformatics online.

The NEPTUN experiments on LOCA thermal-hydraulics for tight-lattice PWRs

International Nuclear Information System (INIS)

Dreier, J.; Chawla, R.; Rouge, N.; Yanar, S.

1990-01-01

The NEPTUN test facility at the Paul Scherrer Institute is currently being used to provide a broad data base for the validation of thermal-hydraulics codes used in predicting the reflooding behaviour of a tight-lattice PWR (light water highb conversion reactor, LWHCR). The present paper gives a description of the facility and the matrix to be covered in the experimental program. Results are presented from a number of forced-feed, bottom-reflooding experiments, comparisons being made with (a) measurements carried out earlier for standard-PWR geometry and (b) the results of a calculational benchmark exercise conducted in the framework of a Swiss/German LWHCR-development agreement. Rewetting for the tight, hexagonal-geometry (p/d = 1.13) NEPTUN-III test bundle has been found to occur in all tests carried out to date, in which reasonably LWHCR-representative values for the various thermal-hydraulics parameters are used. Results of the calculational benchmark exercise have confirmed the need for further code development efforts for achieving reliable predictions of LWHCR reflooding behaviour. (author) 11 figs., 3 tabs., 3 refs

Heat transfer to a dispersed two phase flow and detailed quench front velocity research

International Nuclear Information System (INIS)

De Boer, T.C.; Van der Molen, S.B.

1985-01-01

During the blow-down phase of a loss-off coolant accident (LOCA) in a pressurized water reactor the core will heat up dramatically. Water will be injected in the system, and by bottom flooding the core will be cooled. The use of one-dimensional computer models for the calculation of the reflood process in a bundle needs a better justification. The influence of an unheated shroud on prequench heat transfer is investigated in a tube, an annulus and a 4 rod bundle. By using a glass shroud for the annulus, optical analysis of the dispersed two-phase flow regime has been performed. The ECN 36-rod bundle tests as performed with axial uniform power profile are reflood and boil-down at 0.2 MPa pressure executed for different conditions. The experiment yield a data base suitable for code validation and development. Better understanding is obtained for the influence of the radial non-uniform temperature and/or power distributions on the reflood process. Heat transfer improvement induced by the presence of spacer grids is observed. 72 refs.; 220 figs.

Project of the ATLAS experiment by LHC of CERN

International Nuclear Information System (INIS)

Andrieux, M.L.; Belhorma, B.; Collot, J.; Saintignon, P. de; Dzahini, D.; Ferrari, A.; Hostachy, J.Y.; Martin, Ph.; Rey-Campagnolle, M.; Belymam, A.; Wielers, B.

1997-01-01

The group is involved in the construction of the liquid argon calorimeter of the ATLAS detector. Following an intense R and D phase, the final detailed design at the ATLAS calorimeter was finalized, written and approved by the LHC committee. ATLAS is now in a pre-construction phase which implies that the group activities are mainly devoted to the installation of the assembly line of the electromagnetic pre-sampler sectors. Our R and D activities on the calorimeter electronics were pursued along two lines: the optimization of the filtering amplifiers and a participation to the development of optical links for data transmission. Liquid argon pollution tests under radiation were also achieved at SARA. They proved the radiation hardness of the liquid argon calorimeter. We recently showed that the search for heavy right-handed neutrinos up to m N < 3 TeV is possible with the ATLAS detector. (authors)

Simulation of the QUENCH-06 experiment with MELCOR 1.8.5

International Nuclear Information System (INIS)

Stanojevic, M.; Leskovar, M.

2001-01-01

The MELCOR 1.8.5 code input model and simulation results of the OECD/NEA international standard problem No. 45 (ISP-45) are presented. ISP-45 was performed as QUENCH-06 experiment at Forschungszentrum Karlsruhe. The objectives of the QUENCH program are the analysis of the heat-up, oxidation and delayed reflood phases of a PWR type fuel rod bundle in the QUENCH facility and investigation of the thermal, mechanical, physical and chemical behavior of fuel rod claddings under transient cool-down conditions. The objectives of the OECD/NEA ISP program are the extension of the reflood database to identify key phenomena occurring during accident management measure procedures and code validation, i.e., reliability and accuracy of severe accident codes especially during the quench phase. The QUENCH test bundle is made up of 21 fuel rod simulators approximately 2.5 m long. The Zircaloy-4 rod cladding is identical to that used in pressurized water reactors with respect to material and dimensions. The bundle is heated electrically. The QUENCH-06 experiment had three phases: the pre-oxidation phase, the power transient phase and the reflood-quench phase. According to the ISP-45 specification, the MELCOR 1.8.5 simulation includes the events from the beginning of the pre-oxidation phase until the end of the reflood-quench phase and shut-off of electric power, steam and cooling water. Calculated results are discussed with respect to accuracy, plausibility and completeness. Shortcomings and limitations of the input model are described.(author)

Perlindungan Hukum Kelompok Teisme dalam Sistem Negara Hukum Pancasila

Directory of Open Access Journals (Sweden)

Tomy Michael

2017-05-01

Full Text Available Di era global saat ini, suatu bangsa dituntut mampu bersaing dengan negara lain. Agar tidak terlepas dari unsur khas Indonesia maka penguatan Pancasila sebagai ideologi adalah keharusan. Pancasila yang melingkupi keragaman suku, agama, ras dan antar golongan (SARA kurang tercermin dalam UU Nomor 23 Tahun 2006. Permasalahan yang timbul yaitu hilangnya unsur khas Indonesia yaitu kepercayaan atau agama tradisional karena adanya diskriminasi dengan pengosongan kolom agama dalam Kartu Kelurga (KK dan Kartu Tanda Penduduk (KTP (Pasal 61 dan Pasal 64 UU No. 23-2006. Penelitian ini bertujuan untuk menganalisis korelasi hukum UU No. 23 tahun 2016 dengan keberadaan Pancasila dan SARA di Indonesia. Metode penelitian yang digunakan adalah penelitian hukum normatif. Hasil penelitian ini menunjukkan bahwa UU No. 23 Tahun 2006 tidak selaras dengan semangat Pancasila yang mengakui keragaman SARA dan bertentang dengan asas keadilan dalam UU No. 12Tahun 2011. In the global era, a nation supposedly able to compete with other countries. In order not to be separated from the typical elements of Indonesia, the strengthening of Pancasila as an ideology is a must. Pancasila surrounding ethnic, religious, racial and sectarian (SARA less diversity reflected in Law No. 23 of 2006. The problems that arise, namely the loss of the typical elements of Indonesia namely traditional religious beliefs or because of their discrimination by emptying the religion column in Family Card (KK and Identity Card (KTP (Article 61 and Article 64 of Law No. 23-2006. This study aims to analyze the correlation on Law No. 23of 2016 with the existence of Pancasila and SARA in Indonesia. The method used is a normative legal research. The result of this study reveals that the Law No. 23 of 2006 not in line with the spirit of Pancasila that recognizes the diversity of SARA and incompatible with the principles of justice on Law No. 12 of 2011.

Journal of Food Technology in Africa - Vol 7, No 3 (2002)

African Journals Online (AJOL)

Identification of Lactic Acid Bacteria isolated from Opaque beer (Chibuku) for potential use as a starter culture. Chamunorwa A Togo, Sara B Sara B. Feresu, Anthony Mutukumira. http://dx.doi.org/10.4314/jfta.v7i3.19239 · Performance of different strains of Pleurotus species under Ghanaian conditions. K Vowotor, M Obodai.

A study of the large break loss-of-coolant accident in the Angra-1 nuclear power plant

International Nuclear Information System (INIS)

Borges, E.M.

1984-01-01

The simulation of the Angra-I nuclear power plant under the condition of large break loss of coolant accident is presented, the thermal-hydraulic analysis of the primary circuit during each phase of the acident and thermal analysis of the hottest fuel rod curing reflooding are shown. Computer codes RELAP4/MOD5 (options EM and FLOOD) and TOODEE 2 are used to perform these computations. Fuel rod peak temperatures reached during the simulation are below the permissible levels. However, during the reflooding phase; the maximum oxidation of the cladding exceeds the limit of 0.17 times the original cladding thickness. (Author) [pt

Compartmentalized safety coolant injection system

International Nuclear Information System (INIS)

Johnson, F.T.

1983-01-01

A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

Book Review: Commemoration as Conflict: Space, Memory and Identity in Peace Processes

DEFF Research Database (Denmark)

McQuaid, Sara Dybris

2016-01-01

Book rewiew: Commemoration as Conflict: Space, Memory and Identity in Peace Processes / Sara McDowell and Maire Braniff Palgrave Macmillan, 2014, 224pp., ISBN 978-0-230-27375-7......Book rewiew: Commemoration as Conflict: Space, Memory and Identity in Peace Processes / Sara McDowell and Maire Braniff Palgrave Macmillan, 2014, 224pp., ISBN 978-0-230-27375-7...

Temporal dynamics of in-situ fiber-adherent bacterial community under ruminal acidotic conditions determined by 16S rRNA gene profiling.

Directory of Open Access Journals (Sweden)

Renee M Petri

Full Text Available Subacute rumen acidotic (SARA conditions are a consequence of high grain feeding. Recent work has shown that the pattern of grain feeding can significantly impact the rumen epimural microbiota. In a continuation of these works, the objective of this study was to determine the role of grain feeding patterns on the colonization and associated changes in predicted functional properties of the fiber-adherent microbial community over a 48 h period. Eight rumen-cannulated Holstein cows were randomly assigned to interrupted or continuous 60%-grain challenge model (n = 4 per model to induce SARA conditions. Cows in the continuous model were challenged for 4 weeks, whereas cows of interrupted model had a 1-wk break in between challenges. To determine dynamics of rumen fiber-adherent microbial community we incubated the same hay from the diet samples for 24 and 48 h in situ during the baseline (no grain fed, week 1 and 4 of the continuous grain feeding model as well as during the week 1 following the break in the interrupted model. Microbial DNA was extracted and 16SrRNA amplicon (V3-V5 region sequencing was done with the Illumina MiSeq platform. A significant decrease (P 0.1% relative abundance in the rumen, 18 of which were significantly impacted by the feeding challenge model. Correlation analysis of the significant OTUs to rumen pH as an indicator of SARA showed genus Succiniclasticum had a positive correlation to SARA conditions regardless of treatment. Predictive analysis of functional microbial properties suggested that the glyoxylate/dicarboxylate pathway was increased in response to SARA conditions, decreased between 24h to 48h of incubation, negatively correlated with propanoate metabolism and positively correlated to members of the Veillonellaceae family including Succiniclasticum spp. This may indicate an adaptive response in bacterial metabolism under SARA conditions. This research clearly indicates that changes to the colonizing fiber

Assessing nitrification and denitrification in a paddy soil with different water dynamics and applied liquid cattle waste using the {sup 15}N isotopic technique

Energy Technology Data Exchange (ETDEWEB)

Zhou, Sheng, E-mail: jszs@cc.tuat.ac.jp [Institute of Engineering, Tokyo University of Agriculture and Technology, 2-24-16 Naka-cho, Koganei, Tokyo 184-8588 (Japan); Eco-environmental Protection Institute, Shanghai Academy of Agricultural Sciences, 1000 Jinqi Road, Shanghai 201403 (China); Sakiyama, Yukina; Riya, Shohei [Graduate School of Engineering, Tokyo University of Agriculture and Technology, 2-24-16 Naka-cho, Koganei, Tokyo 184-8588 (Japan); Song, Xiangfu [Eco-environmental Protection Institute, Shanghai Academy of Agricultural Sciences, 1000 Jinqi Road, Shanghai 201403 (China); Terada, Akihiko; Hosomi, Masaaki [Institute of Engineering, Tokyo University of Agriculture and Technology, 2-24-16 Naka-cho, Koganei, Tokyo 184-8588 (Japan)

2012-07-15

Using livestock wastewater for rice production in paddy fields can remove nitrogen and supplement the use of chemical fertilizers. However, paddy fields have complicated water dynamics owing to varying characteristics and would influence nitrogen removal through nitrification followed by denitrification. Quantification of nitrification and denitrification is of great importance in assessing the influence of water dynamics on nitrogen removal in paddy fields. In this study, nitrification and nitrate reduction rates with different water dynamics after liquid cattle waste application were evaluated, and the in situ denitrification rate was determined directly using the {sup 15}N isotopic technique in a laboratory experiment. A significant linear regression correlation between nitrification and the nitrate reduction rate was observed and showed different regression coefficients under different water dynamics. The regression coefficient in the continuously flooded paddy soil was higher than in the drained-reflooded paddy soil, suggesting that nitrate would be consumed faster in the flooded paddy soil. However, nitrification was limited and the maximum rate was only 13.3 {mu}g N g{sup -1} day{sup -1} in the flooded paddy soil with rice plants, which limited the supply of nitrate. In contrast, the drained-reflooded paddy soil had an enhanced nitrification rate up to 56.8 {mu}g N g{sup -1} day{sup -1}, which was four times higher than the flooded paddy soil and further stimulated nitrate reduction rates. Correspondingly, the in situ denitrification rates determined directly in the drained-reflooded paddy soil ranged from 5 to 1035 mg N m{sup -2} day{sup -1}, which was higher than the continuously flooded paddy soil (from 5 to 318 mg N m{sup -2} day{sup -1}) during the vegetation period. The nitrogen removal through denitrification accounted for 38.9% and 9.9% of applied nitrogen in the drained-reflooded paddy soil and continuously flooded paddy soil, respectively

Test facility for rewetting experiments at CDTN

International Nuclear Information System (INIS)

Rezende, Hugo C.; Mesquita, Amir Z.; Ladeira, Luiz C.D.; Santos, Andre A.C.

2015-01-01

One of the most important subjects in nuclear reactor safety analysis is the reactor core rewetting after a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor LWR. Several codes for the prediction of the rewetting evolution are under development based on experimental results. In a Pressurized Water Reactor (PWR) the reflooding phase of a LOCA is when the fuel rods are rewetted from the bottom of the core to its top after having been totally uncovered and dried out. Out-of-pile reflooding experiments performed with electrical heated fuel rod simulators show different quench behavior depending the rods geometry. A test facility for rewetting experiments (ITR - Instalacao de Testes de Remolhamento) has been constructed at the Thermal Hydraulics Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigations on basic phenomena that occur during the reflood phase of a LOCA in a PWR, using tubular and annular test sections. This paper presents the design aspects of the facility, and the current stage of the works. The mechanical aspects of the installation as its instrumentation are described. Two typical tests are presented and results compered with theoretical calculations using computer code. (author)

Assessment of core damage models in SCDAP/RELAP5 during OECD LOFT LP-FP-2

International Nuclear Information System (INIS)

Coryell, E.W.

1991-01-01

The US Nuclear Regulatory Commission has sponsored a program to apply the SCDAP/RELAP5 code to analysis of the transient and reflood phases of the OECD LOFT LP-FP-2 Experiment. The principal objectives of the LP-FP-2 experiment were to determine the fission product release from the fuel during the early phases of a severe fuel damage scenario and to examine the phenomena controlling fission product transport in a vapor/aerosol environment. Calculations with the SCDAP/RELAP5 code, developed at the INEL with NRC support, have been performed to (1) examine the phenomena controlling the progression of both transient and reflood phases of the experiment, (2) enhance our understanding of the phenomena occurring during reflood and add credence to the postulated phenomenological sequence, (3) assess the ability of SCDAP/RELAP5 to examine severe fuel damage issues and phenomena, and (4) identify code strengths and deficiencies with the intent of prioritizing code improvements. Results indicate that the code is able to analyze the early phases of severe fuel damage reasonably well, with potential deficiencies in modelling interaction between molten control rod material and intact fuel

Test facility for rewetting experiments at CDTN

Energy Technology Data Exchange (ETDEWEB)

Rezende, Hugo C.; Mesquita, Amir Z.; Ladeira, Luiz C.D.; Santos, Andre A.C., E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

2015-07-01

One of the most important subjects in nuclear reactor safety analysis is the reactor core rewetting after a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor LWR. Several codes for the prediction of the rewetting evolution are under development based on experimental results. In a Pressurized Water Reactor (PWR) the reflooding phase of a LOCA is when the fuel rods are rewetted from the bottom of the core to its top after having been totally uncovered and dried out. Out-of-pile reflooding experiments performed with electrical heated fuel rod simulators show different quench behavior depending the rods geometry. A test facility for rewetting experiments (ITR - Instalacao de Testes de Remolhamento) has been constructed at the Thermal Hydraulics Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigations on basic phenomena that occur during the reflood phase of a LOCA in a PWR, using tubular and annular test sections. This paper presents the design aspects of the facility, and the current stage of the works. The mechanical aspects of the installation as its instrumentation are described. Two typical tests are presented and results compered with theoretical calculations using computer code. (author)

Accomplishments of LOCA/ECCS experimental research at Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Tasaka, Kanji; Murao, Yoshio; Koizumi, Yasuo

1984-01-01

Japan Atomic Energy Research Institute has investigated loss-of-coolant accident (LOCA)/emergency core cooling system (ECCS) from 1970. Major results of the LOCA/ECCS research are summarized in this report. ROSA-II program was LOCA/ECCS research for a pressurized water reactor (PWR) and ROSA-III program was for a boiling water reactor (BWR). The both test facilities were scaled at approximately 1/400 of the respective reference PWR and BWR. Large scale reflood test is research on reflood phenomena during a large break LOCA of PWR. The test facility is scaled at approximately 1/20 of the reference PWR and the research is still being continued. (author)

Posttest TRAC-PD2/MOD1 predictions for FLECHT SEASET test 31504

International Nuclear Information System (INIS)

Booker, C.P.

1982-01-01

TRAC-PD2/MOD1 is a publicly released version of TRAC that is used primarily to analyze large-break loss-of-coolant accidents in pressurized-water reactors (PWRs). TRAC-PD2 can calculate, among other things, reflood phenomena. TRAC posttest predictions are compared with test 31504 reflood data from the Full-Length Emergency Core Heat Transfer (FLECHT) System Effects and Separate Effects Tests (SEASET) facility. A false top-down quench is predicted near the top of the core and the subcooling is underpredicted at the bottom of the core. However, the overall TRAC predictions are good, especially near the center of the core

GC-MS analysis of the ruminal metabolome response to thiamine supplementation during high grain feeding in dairy cows.

Science.gov (United States)

Xue, Fuguang; Pan, Xiaohua; Jiang, Linshu; Guo, Yuming; Xiong, Benhai

2018-01-01

Thiamine is known to attenuate high-concentrate diet induced subacute ruminal acidosis (SARA) in dairy cows, however, the underlying mechanisms remain unclear. The major objective of this study was to investigate the metabolic mechanisms of thiamine supplementation on high-concentrate diet induced SARA. Six multiparous, rumen-fistulated Holstein cows were used in a replicated 3 × 3 Latin square design. The treatments included a control diet (CON; 20% starch, dry matter basis), a SARA-inducing diet (SAID; 33.2% starch, dry matter basis) and SARA-inducing diet supplemented with 180 mg of thiamine/kg of dry matter intake (SAID + T). On d21 of each period, ruminal fluid samples were collected at 3 h post feeding, and GC/MS was used to analyze rumen fluid samples. PCA and OPLS-DA analysis demonstrated that the ruminal metabolite profile were different in three treatments. Compared with CON treatment, SAID feeding significantly decreased rumen pH, acetate, succinic acid, increased propionate, pyruvate, lactate, glycine and biogenic amines including spermidine and putrescine. Thiamine supplementation significantly decreased rumen content of propionate, pyruvate, lactate, glycine and spermidine; increase rumen pH, acetate and some medium-chain fatty acids. The enrichment analysis of different metabolites indicated that thiamine supplementation mainly affected carbohydrates, amino acids, pyruvate and thiamine metabolism compared with SAID treatment. These findings revealed that thiamine supplementation could attenuate high-concentrate diet induced SARA by increasing pyruvate formate-lyase activity to promote pyruvate to generate acetyl-CoA and inhibit lactate generation. Besides, thiamine reduced biogenic amines to alleviate ruminal epithelial inflammatory response.

Seismic Amplitude Ratio Analysis of the 2014-2015 Bár∂arbunga-Holuhraun Dike Propagation and Eruption

Science.gov (United States)

Caudron, Corentin; White, Robert S.; Green, Robert G.; Woods, Jennifer; Ágústsdóttir, Thorbjörg; Donaldson, Clare; Greenfield, Tim; Rivalta, Eleonora; Brandsdóttir, Bryndís.

2018-01-01

Magma is transported in brittle rock through dikes and sills. This movement may be accompanied by the release of seismic energy that can be tracked from the Earth's surface. Locating dikes and deciphering their dynamics is therefore of prime importance in understanding and potentially forecasting volcanic eruptions. The Seismic Amplitude Ratio Analysis (SARA) method aims to track melt propagation using the amplitudes recorded across a seismic network without picking the arrival times of individual earthquake phases. This study validates this methodology by comparing SARA locations (filtered between 2 and 16 Hz) with the earthquake locations (same frequency band) recorded during the 2014-2015 Bár∂arbunga-Holuhraun dike intrusion and eruption in Iceland. Integrating both approaches also provides the opportunity to investigate the spatiotemporal characteristics of magma migration during the dike intrusion and ensuing eruption. During the intrusion SARA locations correspond remarkably well to the locations of earthquakes. Several exceptions are, however, observed. (1) A low-frequency signal was possibly associated with a subglacial eruption on 23 August. (2) A systematic retreat of the seismicity was also observed to the back of each active segment during stalled phases and was associated with a larger spatial extent of the seismic energy source. This behavior may be controlled by the dike's shape and/or by dike inflation. (3) During the eruption SARA locations consistently focused at the eruptive site. (4) Tremor-rich signal close to ice cauldrons occurred on 3 September. This study demonstrates the power of the SARA methodology, provided robust site amplification; Quality Factors and seismic velocities are available.

Usaha La Sangkuru Patau Dalam Mengembangkan Agama Islam Di Kerajaan Wajo

Directory of Open Access Journals (Sweden)

Nasruddin Nasruddin

2014-12-01

Full Text Available La Sangkuru Patau is one of the Arung Matoa Wajo has substantial powers. His remarks could be a law and society Wajo stick to what he says. So when La Sangkuru Patau said that he had converted to Islam, the community participated Wajo also embraced Islam. Arung Matoa Wajo La Sangkuru with doing business in developing Islam in Wajo that bring Datu Sulaiman to provide Islamic religious instruction in public Wajo, then set up a sara officers' special charge of handling issues such royal Friday prayers, warnings Islamic holidays , then hold the officer cadre sara 'that will be placed outside the capital of the kingdom (Tosora, sara officers' given the facilities, such as free of taxes, as well as the parallel position to the position of indigenous stakeholders.

Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

International Nuclear Information System (INIS)

Fletcher, C.D.; Callow, R.A.

1989-01-01

The USNRC recently identified a possible safety concern for PWR's. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: 1) describes the important aspects of the problem, 2) provides an initial analysis of the consequences, and 3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of WE design, the concern is greatest for those plants. There is less concern for most plants of CE design, and likely no concern for plants of BW design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a WE PWR. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed. (orig./HP)

Assessing the viability of the Species at Risk Act in managing commercial exploitation and recovery of threatened and endangered marine fish in Canada

OpenAIRE

Druce, Courtney Danielle

2012-01-01

Commercially exploited threatened or endangered marine fish are consistently declined for listing under Canada’s Species at Risk Act (SARA), largely due to predicted socio-economic impacts associated with SARA’s prohibitions. However, commercial exploitation can be exempted from SARA’s general prohibitions. If exemptions were utilized, commercially exploited species could benefit from other aspects of SARA listing, and support continued economic opportunities for fishers. I conducted a litera...

ISP42 (PANDA Tests) - Blind Phase Comparison Report

International Nuclear Information System (INIS)

Luebbesmeyer, D.; Aksan, S.N.

2003-05-01

The International Standard Problem (ISP) No. 45 is part of the overall ISP program of the OECD/NEA and is dedicated to the behavior of heat-up and delayed reflood of fuel elements in nuclear reactors during a hypothetical accident. ISP-45 is related to the out-of-pile bundle quench experiment QUENCH-06, performed at Forschungszentrum Karlsruhe (FZK), Germany, on December 13, 2000. Special attention was paid to hydrogen production. To assess the ability of severe accident codes to simulate processes during core heat-up and reflood at temperatures above 2000 K, the behavior of the bundle during the whole experiment should be calculated on the basis of the necessary experimental initial and boundary conditions, but without knowing further experimental details. In this so-called blind phase 21 participants from 15 nations contributed with 8 different code systems (ATHLET-CD, ICARE/CATHARE, IMPACT/SAMPSON, GENFLO, MAAP, MELCOR, SCDAPSIM, SCDAP-3D). Additionally, posttest calculations using the in-house version SCDAP/RELAP5 mod3.2.irs are used for comparison. After the end of the blind phase all measured data were made available and the participants were invited to deliver a second calculation, where this knowledge could be used (so called open phase). In this report, results of the blind calculations are presented, analyzed, and compared to experimental data. During heat-up most results do not deviate significantly from one another, except as a consequence of some obvious user errors, so that a definition of a mainstream is justified. For the quench phase the lack of adequate hydraulic modeling becomes obvious: some participants could not match the observed cool-down rates, others had to use very fine meshes to compensate code deficiencies. To overcome this insufficiency some newly developed reflood models were used in MAAP and MELCOR. In QUENCH-06, oxide layers were thick enough to protect the cladding from melting and failure below 2200 K, so that no massive hydrogen

Comparison and Interpretation Report of the OECD International Standard Problem No. 45 - Exercise (QUENCH-06)

International Nuclear Information System (INIS)

Hering, W.; Homann, Ch.; Lamy, J.S.; Miassoedov, A.; Schanz, G.; Sepold, L.; Steinbrueck, M.

2002-10-01

The International Standard Problem (ISP) No. 45 is part of the overall ISP program of the OECD/NEA and is dedicated to the behavior of heat-up and delayed reflood of fuel elements in nuclear reactors during a hypothetical accident. ISP-45 is related to the out-of-pile bundle quench experiment QUENCH-06, performed at Forschungszentrum Karlsruhe (FZK), Germany, on December 13, 2000. Special attention was paid to hydrogen production. To assess the ability of severe accident codes to simulate processes during core heat-up and reflood at temperatures above 2000 K, the behavior of the bundle during the whole experiment should be calculated on the basis of the necessary experimental initial and boundary conditions, but without knowing further experimental details. In this so-called blind phase 21 participants from 15 nations contributed with 8 different code systems (ATHLET-CD, ICARE/CATHARE, IMPACT/SAMPSON, GENFLO, MAAP, MELCOR, SCDAPSIM, SCDAP-3D). Additionally, posttest calculations using the in-house version SCDAP/RELAP5 mod3.2.irs are used for comparison. After the end of the blind phase all measured data were made available and the participants were invited to deliver a second calculation, where this knowledge could be used (so-called open phase). In this report, results of the blind calculations are presented, analyzed, and compared to experimental data. During heat-up most results do not deviate significantly from one another, except as a consequence of some obvious user errors, so that a definition of a mainstream is justified. For the quench phase the lack of adequate hydraulic modeling becomes obvious: some participants could not match the observed cool-down rates, others had to use very fine meshes to compensate code deficiencies. To overcome this insufficiency some newly developed reflood models were used in MAAP and MELCOR. In QUENCH-06, oxide layers were thick enough to protect the cladding from melting and failure below 2200 K, so that no massive hydrogen

Validation of ASTECV2.1 based on the QUENCH-08 experiment

Energy Technology Data Exchange (ETDEWEB)

Gómez-García-Toraño, Ignacio, E-mail: ignacio.torano@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology (INR), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Sánchez-Espinoza, Víctor-Hugo; Stieglitz, Robert [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology (INR), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Stuckert, Juri [Karlsruhe Institute of Technology, Institute for Applied Materials-Applied Materials Physics (IAM-AWP), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Laborde, Laurent; Belon, Sébastien [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Nuclear Safety Division/Safety Research/Severe Accident Department, Saint Paul Lez Durance 13115 (France)

2017-04-01

Highlights: • ASTECV2.1 can reproduce QUENCH-08 experimental trends e.g. hydrogen generation. • Radial temperature gradient and heat transfer through argon gap are underestimated. • Mesh sizes lower than 55 mm needed to capture the strong axial temperature gradient. • Minor variations of external electrical resistance strongly affect bundle heat-up. • Modelling of a bypass and inclusion of currents partially overcome discrepancies. - Abstract: The Fukushima accidents have shown that further improvements of Severe Accident Management Guidelines (SAMGs) are still necessary. Hence, the enhancement of severe accident codes and their validation based on integral experiments is pursued worldwide. In particular, the capabilities of the European integral severe accident ASTECV2.1 code are being extended within the CESAM project through the improvement of physical models, code numerics and an extensive code validation. Among the different strategies encompassed in the plant SAMGs, one of the most important ones to prevent core damage is the injection of water into the overheated core (reflooding). However, under certain conditions, reflooding may trigger a sharp hydrogen generation that may jeopardize the containment. Within this work, ASTECV2.1 models describing the early in-vessel phase of the severe accident and its termination by core reflooding are validated against data from the QUENCH test facility. The QUENCH-08, involving the injection of 15 g/s (about 0.6 g/s/rod) of saturated steam at a bundle temperature of 2073 K, has been selected for this comparison. Results show that ASTECV2.1 is able to reproduce the experimental temperatures and oxide thicknesses at representative bundle locations. The predicted total hydrogen generation (76 g) is similar to the experimental one (84 g). In addition, the choices of an axial mesh size lower than 55 mm and of an external electrical resistance of a 7 mΩ/rod have been justified with parametric analyses. Finally, new

A mechanistic Eulerian-Lagrangian model for dispersed flow film boiling

International Nuclear Information System (INIS)

Andreani, M.; Yadigaroglu, G.

1991-01-01

In this paper a new mechanistic model of heat transfer in the dispersed flow regime is presented. The usual assumptions that render most of the available models unsuitable for the analysis of the reflooding phase of the LOCA are discussed, and a two-dimensional time-independent numerical model is developed. The gas temperature field is solved in a fixed-grid (Eulerian) mesh, with the droplets behaving as mass and energy sources. The histories of a large number of computational droplets are followed in a Lagrangian frame, considering evaporation, break-up and interactions with the vapor and with the wall. comparisons of calculated wall and vapor temperatures with experimental data are shown for two reflooding tests

Disparities in availability of essential medicines to treat non-communicable diseases in Uganda: A Poisson analysis using the Service Availability and Readiness Assessment.

Science.gov (United States)

Armstrong-Hough, Mari; Kishore, Sandeep P; Byakika, Sarah; Mutungi, Gerald; Nunez-Smith, Marcella; Schwartz, Jeremy I

2018-01-01

Although the WHO-developed Service Availability and Readiness Assessment (SARA) tool is a comprehensive and widely applied survey of health facility preparedness, SARA data have not previously been used to model predictors of readiness. We sought to demonstrate that SARA data can be used to model availability of essential medicines for treating non-communicable diseases (EM-NCD). We fit a Poisson regression model using 2013 SARA data from 196 Ugandan health facilities. The outcome was total number of different EM-NCD available. Basic amenities, equipment, region, health facility type, managing authority, NCD diagnostic capacity, and range of HIV services were tested as predictor variables. In multivariate models, we found significant associations between EM-NCD availability and region, managing authority, facility type, and range of HIV services. For-profit facilities' EM-NCD counts were 98% higher than public facilities (p < .001). General hospitals and referral health centers had 98% (p = .004) and 105% (p = .002) higher counts compared to primary health centers. Facilities in the North and East had significantly lower counts than those in the capital region (p = 0.015; p = 0.003). Offering HIV care was associated with 35% lower EM-NCD counts (p = 0.006). Offering HIV counseling and testing was associated with 57% higher counts (p = 0.048). We identified multiple within-country disparities in availability of EM-NCD in Uganda. Our findings can be used to identify gaps and guide distribution of limited resources. While the primary purpose of SARA is to assess and monitor health services readiness, we show that it can also be an important resource for answering complex research and policy questions requiring multivariate analysis.

Systems analysis programs for hands-on integrated reliability evaluations (SAPHIRE) version 5.0

International Nuclear Information System (INIS)

Russell, K.D.; Kvarfordt, K.J.; Skinner, N.L.; Wood, S.T.

1994-07-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. This volume is the reference manual for the Systems Analysis and Risk Assessment (SARA) System Version 5.0, a microcomputer-based system used to analyze the safety issues of a open-quotes familyclose quotes [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. The SARA database contains PRA data primarily for the dominant accident sequences of a family and descriptive information about the family including event trees, fault trees, and system model diagrams. The number of facility databases that can be accessed is limited only by the amount of disk storage available. To simulate changes to family systems, SARA users change the failure rates of initiating and basic events and/or modify the structure of the cut sets that make up the event trees, fault trees, and systems. The user then evaluates the effects of these changes through the recalculation of the resultant accident sequence probabilities and importance measures. The results are displayed in tables and graphs that may be printed for reports. A preliminary version of the SARA program was completed in August 1985 and has undergone several updates in response to user suggestions and to maintain compatibility with the other SAPHIRE programs. Version 5.0 of SARA provides the same capability as earlier versions and adds the ability to process unlimited cut sets; display fire, flood, and seismic data; and perform more powerful cut set editing

Service life prediction. Development of models for predicting the service life of power plant components subject to thermomechanical creep fatigue; Lebensdauervorhersage. Entwicklung von Modellen zur Lebensdauervorhersage von Kraftwerksbauteilen unter thermisch-mechanischer Kriechermuedungsbeanspruchung. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Cui, L.; Scholz, A. [Technische Univ. Darmstadt (Germany). Institut fuer Werkstoffkunde; Hartrott, P. von; Schlesinger, M. [Fraunhofer-Institut fuer Werkstoffmechanik (IWM), Freiburg im Breisgau (Germany)

2009-07-01

Extensive use is made of massive components of heat resistant and highly heat resistant materials in installations of the power and heating industry. These components are exposed to varying thermomechanical stress as a result of ramping-up and down processes. In this research project two computer-assisted methods of predicting service life until crack initiation were extended to include cases of thermomechanical multi-axis stress conducive to creep fatigue and of superposition of high-cycle stress on power plant components. Investigations were limited to rotor steel of type X12CrMoWVNbN10-1-1. Complex thermomechanical multi-axis experiments were performed on round, notched and cruciform test specimens of close-to-life dimensions in order to demonstrate by experiment the validity of these models. The results of these calculations showed an acceptable degree of agreement between experiment and simulation for both models. Calculations on earlier TMF experiments performed at IfW on hollow specimens of 1%CrMoNiV showed good predictability for both the SARA and the ThoMat programme. Calculations on experiments performed at MPA Stuttgart on model bodies consisting of the same 1%CrMoNiV showed a predictability of acceptable variability considering the complexity of the stresses involved. A further outcome of this project is that the use of SARA appears universally suitable for the construction of new plants and in the service area, while the use of ThoMat appears suited for detail optimisation in the development process.

Ruminal acidosis: a review with detailed reference to the controlling agent Megasphaera elsdenii NCIMB 41125

OpenAIRE

Meissner, H.H.; Henning, P.H.; Horn, C.H.; Leeuw, K-J.; Hagg, F.M.; Fouché, G.

2010-01-01

Ruminal acidosis is discussed with reference to causes and economic and health implications. Distinction is made between the acute form which with proper adaptation to high energy diets is seldom encountered and the more problematic chronic or sub-acute form, commonly referred to as sub-acute ruminal acidosis (SARA). Apart from stepwise transition from roughage to concentrates, methods adopted to reduce SARA include grain treatment to reduce starch degradation, feed additives such as buffers ...

An eHealth Project on Invasive Pneumococcal Disease: Comprehensive Evaluation of a Promotional Campaign.

Science.gov (United States)

Panatto, Donatella; Domnich, Alexander; Gasparini, Roberto; Bonanni, Paolo; Icardi, Giancarlo; Amicizia, Daniela; Arata, Lucia; Carozzo, Stefano; Signori, Alessio; Bechini, Angela; Boccalini, Sara

2016-12-02

. Similarly, the postintervention daily trend in the number of users was positive, with a relative increase of 0.9% (95% CI 0.0%-1.8%) for the app and 1.4% (95% CI 0.7%-2.1%) for the website. Demographics differed between app and website users and Facebook fans. A total of 69.15% (10,793/15,608) of users could be defined as being at risk of IPD, while 4729 users expressed intentions to ask their doctor for further information on IPD. The mean app quality score assigned by end users was approximately 79.5% (397/500). Despite its specific topic, Pneumo Rischio was accessed by a considerable number of users, who ranked it as a high-quality project. In order to reach their target populations, however, such projects should be promoted. ©Donatella Panatto, Alexander Domnich, Roberto Gasparini, Paolo Bonanni, Giancarlo Icardi, Daniela Amicizia, Lucia Arata, Stefano Carozzo, Alessio Signori, Angela Bechini, Sara Boccalini. Originally published in the Journal of Medical Internet Research (http://www.jmir.org), 02.12.2016.

Sharing AIS Related Anomalies (SARA)

Science.gov (United States)

2016-03-01

78 6.3.7 SQL Versus NoSQL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 6.4 Data Processing...43 6.1 Overview of SQL and NoSQL differences, from [56]. . . . . . . . . . . . . . . . . . 82 A.1 Description of the ship anomaly upload use...constraints must be considered. These requirements, however, can only be defined when lower level implementation decisions, such as SQL versus NoSQL

Quench/reflood modeling in MELCOR

International Nuclear Information System (INIS)

Gauntt, R.O.

2001-01-01

The authors describe the reactor accident simulation model MELCOR. It comprises hydrodynamic investigations on reactor core quenching, hydrogen generation in the reactor core vessel, quench front advances. Preliminary comparisons to data are reasonable but need further validation. (uke)

Altered TGF-β endocytic trafficking contributes to the increased signaling in Marfan syndrome.

Science.gov (United States)

Siegert, Anna-Maria; Serra-Peinado, Carla; Gutiérrez-Martínez, Enric; Rodríguez-Pascual, Fernando; Fabregat, Isabel; Egea, Gustavo

2018-02-01

The main cardiovascular alteration in Marfan syndrome (MFS) is the formation of aortic aneurysms in which augmented TGF-β signaling is reported. However, the primary role of TGF-β signaling as a molecular link between the genetic mutation of fibrillin-1 and disease onset is controversial. The compartmentalization of TGF-β endocytic trafficking has been shown to determine a signaling response in which clathrin-dependent internalization leads to TGF-β signal propagation, and caveolin-1 (CAV-1) associated internalization leads to signal abrogation. We here studied the contribution of endocytic trafficking compartmentalization to increased TGF-β signaling in vascular smooth muscle cells (VSMC) from MFS patients. We examined molecular components involved in clathrin- (SARA, SMAD2) and caveolin-1- (SMAD7, SMURF2) dependent endocytosis. Marfan VSMC showed higher recruitment of SARA and SMAD2 to membranes and their increased interaction with TGF-β receptor II, as well as higher colocalization of SARA with the early endosome marker EEA1. We assessed TGF-β internalization using a biotinylated ligand (b-TGF-β), which colocalized equally with either EEA1 or CAV-1 in VSMC from Marfan patients and controls. However, in Marfan cells, colocalization of b-TGF-β with SARA and EEA1 was increased and accompanied by decreased colocalization with CAV-1 at EEA1-positive endosomes. Moreover, Marfan VSMC showed higher transcriptional levels and membrane enrichment of RAB5. Our results indicate that increased RAB5-associated SARA localization to early endosomes facilitates its TGF-β receptor binding and phosphorylation of signaling mediator SMAD2 in Marfan VSMC. This is accompanied by a reduction of TGF-β sorting into multifunctional vesicles containing cargo from both internalization pathways. Copyright © 2017 Elsevier B.V. All rights reserved.

A numerical solution model of the rewetting of a nuclear fuel rod

International Nuclear Information System (INIS)

Braz Filho, F.A.

1984-01-01

The study of thermal behaviour of a nuclear reactor fuel rod during the reflooding phase of the loss-of-coolant accident (LOCA) is presented. A mathematical model and a numerical scheme were proposed in order to solve the bidimensional heat conduction equation in cylindrical coordinates. The phenomenon of reflooding is not completely understood. One of the main difficulties is to estimate the heat transfer coefficient (h). For this reason two different models were elaborated: in the first three regions are considered and in each region h is considered constant; in the second the h profile is adjusted according to the boiling curve. The three region model yields satisfactory results at high and low mass flows while the 'boiling curve' model yields reasonable at low flows. (Author) [pt

Peranan Raja La Maddaremmeng Dalam Penyebaran Islam Di Bone

Directory of Open Access Journals (Sweden)

Nasruddin Nasruddin

2014-06-01

Full Text Available King Bone to Lamaddaremmeng XIII has a very large role and has the effect of early-early development of Islam in southern Sulawesi. He Maddaremeng is committed to enforce Sharia Law. His efforts were Eliminating slavery, Addition parewa Sara, to revamp the governance structure by adding positions previously unknown structure that positions parewa sara, eradication of idols, eliminating the arbitrary actions on others, Equal rights. Other efforts are trying to eliminate the differences in the degree of the community.

Management of Sub-acute Ruminal Acidosis in Dairy Cattle for Improved Production: A Review

OpenAIRE

Kafil Hussain; Amjad Ul Islam; Surinder Kumar Gupta

2011-01-01

Sub-acute ruminal acidosis (SARA) is a well-recognized digestive disorder that is an increasing health problem in most dairy herds. Feeding diets high in grain and other highly fermentable carbohydrates to dairy cows increases milk production, but also increases the risk of SARA. Sub-acute ruminal acidosis is defined as periods of moderately depressed ruminal pH, from about 5.5 to 5.0. Sub-acute ruminal acidosis may be associated with laminitis and other health problems resulting in decreased...

Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

International Nuclear Information System (INIS)

1997-08-01

In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

Analysis of the return to power scenario following a LBLOCA in a PWR

Energy Technology Data Exchange (ETDEWEB)

Macian, R.; Tyler, T.N.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)

1995-09-01

The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus, the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.

Steam generator tube rupture effects on a LOCA

International Nuclear Information System (INIS)

LaChance, J.L.

1979-01-01

A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

Analysis of Semiscale Mod-1 integral test with asymmetrical break (Test S-29-1)

International Nuclear Information System (INIS)

Langerman, M.A.

1977-03-01

Selected experimental data obtained from Semiscale Mod-1 cold leg break Test S-29-1 and results obtained from analytical codes are analyzed. This test was the first integral blowdown reflood test conducted with the Mod-1 system and was a special test designed specifically to evaluate the sensitivity of the early Mod-1 core thermal response (0 to 5 sec after rupture) to the magnitude and direction of the core flow. To achieve this specific objective in Test S-29-1, the vessel side break area was reduced to approximately one-half the scaled break area associated with a 200 percent cold leg break test. The reduction in break area significantly reduced the core flow reversal that took place immediately after rupture and resulted in periods of positive core flow in the early portion of the test. The results obtained from this test are compared with results obtained from a 200 percent cold leg break test and the effect of core flow on early core thermal response is evaluated. Since Test S-29-1 was the first integral blowdown reflood test conducted with the Mod-1 system, data are also presented through the reflood stage of the test and the results are analyzed. The test data and the core thermal response calculated with the RELAP4 code are also compared

SEFLEX - fuel rod simulator effects in flooding experiments. Pt. 2

International Nuclear Information System (INIS)

Ihle, P.; Rust, K.

1986-03-01

This report presents typical data and a limited heat transfer analysis from unblocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5 x 5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5 x 5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

SEFLEX fuel rod simulator effects in flooding experiments. Pt. 3

International Nuclear Information System (INIS)

Ihle, P.; Rust, K.

1986-03-01

This report presents typical data and a limited heat transfer analysis from blocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5x5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5x5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

Loss-of-coolant accident test series TC-1 experiment operating specifications

International Nuclear Information System (INIS)

Yackle, T.R.

1979-09-01

The purpose of this document is to specify the experiment operating procedure for the test series TC-1. The effects of externally mounted cladding thermocouples on the fuel rod thermal behavior during LOCA blowdown and reflood cycles will be investigated in the test. Potential thermocouple effects include: (a) delayed DNB, (b) momentary cladding rewets following DNB, (c) premature cladding rewet during a blowdown two-phase slug period, and (d) early cladding rewet during reflood. The two-phase slug period will be controlled by momentarily opening the hot leg valve. The slug will consist of lower plenum liquid that is sent through the flow shrouds and will be designed to quench the fuel rods at a rate that is similar to the slug experienced early in the LOFT L2-2 and L2-3 tests

Fuel rod simulator effects in flooding experiments single rod tests

International Nuclear Information System (INIS)

Nishida, M.

1984-09-01

The influence of a gas filled gap between cladding and pellet on the quenching behavior of a PWR fuel rod during the reflood phase of a LOCA has been investigated. Flooding experiments were conducted with a short length electrically heated single fuel rod simulator surrounded by glass housing. The gap of 0.05 mm width between the Zircaloy cladding and the internal Al 2 O 3 pellets of the rod was filled either wit helium or with argon to vary the radial heat resistance across the gap. This report presents some typical data and an evaluation of the reflood behavior of the fuel rod simulator used. The results show that the quench front propagates faster for increasing heat resistance in the gap between cladding and heat source of the rod. (orig.) [de

Oil and gas property transfers: Analyzing the environmental risk through the environmental site assessment process

International Nuclear Information System (INIS)

Bratberg, D.; Hocker, S.

1994-01-01

The Superfund Act made anyone buying contaminated real estate liable for cleanup costs whether they know about the contamination or contributed to the contamination. In 1986, SARA amended the Superfund Act to include a provision known as the ''Innocent Landowner Defense.'' This provision created a defense for purchasers of contaminated property who did not contribute to the contamination and had no reason to believe that the property was contaminated at the time of the real estate transfer. SARA allows the purchasers and lenders to perform an environmental assessment using ''due diligence'' to identify contamination problems existing at a site. Since the passing of SARA, the environmental site assessment (ESA) process has become commonplace during the transfer of commercial real estate. Since the introduction of SARA, many professional associations, governmental agencies, and proposed federal legislation have struggled to produce a standard for conducting Phase 1 ESAs. Only recently has a standard been produced. Until recently, the domestic oil and gas industry has been relatively unconcerned about the Superfund liability issues. This approach was created by Congress's decision in 1980 to temporarily exempt the majority of oil and gas exploration and production wastes from federal hazardous waste rulings. However, new stringent rules governing oil and gas waste management practices are being considered by federal and state regulatory agencies. Based upon this knowledge and the awakening of public awareness, the use of ESAs for oil and gas transactions is increasing

The concentration principle applied to spaceborne solar arrays. AGORA mission: Studies synthesis

Science.gov (United States)

Laget, R.

1986-01-01

Studies that led to selection of the distributed 25 kW SARA LOUVRE concept for the solar cell generator to be flown on the AGORA asteroid mission, and the major characteristics of such a spaceborne solar array are summarized. In the SARA LOUVRE concept, a parabolic cross section reflector concentrates incident light over the rear face of the identical, preceding reflector dish. The whole set of reflectors is pivotally commanded, thus compensating the effects of depointing. Geometric concentration factor is 10. End of life power level at 2.5 AU is 4.5 kW.

Development of a biofilm inhibitor molecule against multidrug resistant Staphylococcus aureus associated with gestational urinary tract infections

Directory of Open Access Journals (Sweden)

Balamurugan eP

2015-08-01

Full Text Available Urinary Tract Infection (UTI is a globally widespread human infection caused by an infestation of uropathogens. Eventhough, Escherichia coli is often quoted as being the chief among them, Staphylococcus aureus involvement in UTI especially in gestational UTI is often understated. Staphylococcal accessory regulator A (SarA is a quorum regulator of S. aureus that controls the expression of various virulence and biofilm phenotypes. Since SarA had been a focussed target for antibiofilm agent development, the study aims to develop a potential drug molecule targeting the SarA of S. aureus to combat biofilm associated infections in which it is involved. In our previous studies, we have reported the antibiofilm activity of SarA based biofilm inhibitor, (SarABI with a 50% minimum biofilm inhibitory concentration (MBIC50 value of 200 µg/mL against S. aureus associated with vascular graft infections and also the antibiofilm activity of the root ethanolic extracts of Melia dubia against uropathogenic E. coli. In the present study, in silico design of a hybrid molecule composed of a molecule screened from M. dubia root ethanolic extracts and a modified SarA based inhibitor (SarABIM was undertaken. SarABIM is a modified form of SarABI where the fluorine groups are absent in SarABIM. Chemical synthesis of the hybrid molecule, 4-(Benzylaminocyclohexyl 2-hydroxycinnamate (henceforth referred to as UTI Quorum-Quencher, UTIQQ was then performed, followed by in vitro and in vivo validation. The MBIC¬50 and MBIC90 of UTIQQ were found to be 15 µg/mL and 65 µg/mL respectively. Confocal laser scanning microscopy (CLSM images witnessed biofilm reduction and bacterial killing in either UTIQQ or in combined use of antibiotic gentamicin and UTIQQ. Similar results were observed with in vivo studies of experimental UTI in rat model. So, we propose that the drug UTIQQ would be a promising candidate when used alone or, in combination with an antibiotic for staphylococcal

Heat transfer to a dispersed two-phase flow and detailed quench front velocity research

International Nuclear Information System (INIS)

Boer, T.C. de; Molen, S.B. van der

1985-01-01

A programme to obtain a data base for 'Boildown and Reflood' computer code development and to obtain information on the influence of non-uniform temperature and/or power profile on the quench front velocity and prequench heat transfer, including unheated wall and grid effects, has been undertaken. It is in two parts. In the first (for the tube, annulus and a 4-rod bundle) an early wetting of the unheated shroud is shown. This leads to an increase in quench front velocity and in liquid transport downstream from the quench front. For the inverted annular flow regime the extended Bromley correlation gives good agreement with the experimental data. In the second part (36-rod bundle reflood test programme) the wall-temperature differences in the radial direction gives rise to heat transfer processes which are described and explained. (U.K.)

ROSA-II test data report, 10

International Nuclear Information System (INIS)

1977-12-01

Results of the ROSA-II test simulating a loss-of coolant accident (LOCA) in a light water reactor (LWR) are presented, including the test conditions and interpretation of the phenomena for test runs 415, 417, 421 and 422. Even in small break at the cold leg, the core is exposed to void and the temperature rises. In small break of the hot leg, however, core cooling keeps without temperature rise, because there still remains much residual water and upward core flow exists. Direct effect of the HPCI on the depressurization rate is small, but it increases the accumulator injection rate, leading to early core reflooding and early core cooling from upward. Effects of the secondary system depressurization are increase of depressurization and discharge rates of the primary loop, which results in early initiation of the accumulator injection and core reflooding. (auth.)

Experiment data report for Semiscale Mod-1 Test S-05-5 (alternate ECC injection test)

International Nuclear Information System (INIS)

Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

1977-04-01

Recorded test data are presented for Test S-05-5 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-5 was conducted from initial conditions of 2263 psia and 537 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. The upper plenum was vented through a reflood bypass line interconnecting the hot and cold legs of the broken loop

Properties of solar generators with reflectors and radiators

Science.gov (United States)

Ebeling, W. D.; Rex, D.; Bierfischer, U.

1980-06-01

Radiation cooled concentrator systems using silicon and GaAs cells were studied. The principle of radiation cooling by the reflector surfaces is discussed for cylindrical parabolic reflectors (SARA), truncated hexagonal pyramids, and a small trough configuration. Beam paths, collection properties for imperfect orientation, and thermal optimization parameters were analyzed. The three concentrating systems with radiation cooling offer advantages over the plane panel and over the large trough. With silicon solar cells they exhibit considerably lower solar cell consumption per Kw and also lower mass per kW. With GaAs cells the SARA system reduces the number of solar cells needed per kW to less than 10%. Also in all other cases SARA offers the best values for alpha and F sub sol, as long as narrow angular tolerances of the panel orientation can be met. Analysis of the energy collecting properties for imperfect orientation shows the superiority of the hexagonal concentrator. This device can produce power for even large angles between the sun and the panel normal.

Gene : CBRC-SARA-01-0604 [SEVENS

Lifescience Database Archive (English)

Full Text Available and polyadenylation specific factor 1, 160kDa, isoform CRA_a [Homo sapiens] gb|EAW82107.1| cleavage and poly...adenylation specific factor 1, 160kDa, isoform CRA_a [Homo sapiens] gb|EAW82108.1| cleavage and polyadenylat...ion specific factor 1, 160kDa, isoform CRA_a [Homo sapiens] 3e-16 77% MSTVPTVSCTS...TDRWGLVGTGRCYGSGRGGSMLASCGELSQAIALPKPVLSGHQAGSDPAGSSLYLVLPEGRCLPSVPAHVCPGTAEPAQHSQTPLQRELRISVLPAYLSYDAPWPVRKIPLRCTAHYVAYHVESKVCPWAGVPRAGRPGPQAQPALSFRCMQWPPAPIRHAPASHA ...

Pneumonia por varicela associada com síndrome da angústia respiratória aguda: relato de dois casos Varicella pneumonia complicated with acute respiratory distress syndrome: two cases report

OpenAIRE

Marcelo Moreno; Ricardo Castelão; Susana Orrico Peres; Suzana Margareth Lobo

2007-01-01

JUSTIFICATIVA E OBJETIVOS: A varicela é uma doença exantemática causada pela infecção primária do vírus varicela zoster (VVZ). A pneumonia pelo VVZ complicada com a síndrome da angústia respiratória aguda (SARA) é rara e associa-se a altas taxas de morbimortalidade. O objetivo deste estudo foi apresentar dois casos de pneumonia por varicela que evoluíram com SARA e outras disfunções orgânicas. RELATO DOS CASOS: Paciente de 15 anos, imunocomprometido com a síndrome da imunodeficiência adquirid...

Experimental observation of the droplet size change across a wet grid spacer in a 6 × 6 rod bundle

International Nuclear Information System (INIS)

Cho, Hyoung Kyu; Choi, Ki Yong; Cho, Seok; Song, Chul-Hwa

2011-01-01

Highlights: ► In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. ► The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle. ► The major measuring parameters of the experiment were the droplet size and velocity. ► This test provided the data on the change of the droplet size after collision with a wet grid spacer. - Abstract: During the reflood phase of a postulated loss of coolant accident in a nuclear reactor, entrainment of liquid droplets can occur at a quench front of reflooding water. It is widely recognized that the behavior of the entrained droplets crucially affects the reflood heat transfer phenomena by decreasing the superheated steam temperature and interacting with a rod bundle and spacer grids. For this reason, various experimental and numerical studies have been performed to examine droplet behavior such as the droplet size, velocity and droplet fraction inside a rod array. In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle geometry, which results in the change of the interfacial heat transfer between droplets and superheated steam. A 6 × 6 rod bundle test facility in Korea Atomic Energy Research Institute was used for the experiment. Steam was supplied by an external boiler into the bottom of the test channel, and a droplet injection nozzle was equipped instead of simulating a quench front of reflooding water. The major measuring parameters of the experiment were the droplet size and velocity, which were measured by a high-speed camera and a digital image processing technique. A series of experiments were conducted with various flow conditions of a steam injection velocity, heater temperature, droplet size, and droplet flow rate. The experiments provided the data on the change of the Sauter mean diameter of

CFD analysis of blockage length on a partially blocked fuel rod

International Nuclear Information System (INIS)

Scuro, Nikolas Lymberis; Andrade, Delvonei Alves de; Angelo, Gabriel; Angelo, Edvaldo

2017-01-01

In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod. (author)

SCTF Core-I test results

International Nuclear Information System (INIS)

Adachi, Hiromichi; Sudo, Yukio; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Hirano, Kemmei

1982-07-01

The Slab Core Test Facility (SCTF) of Japan Atomic Energy Research Institute (JAERI) was constructed to investigate two-dimensional thermohydrodynamics in the core and the communication in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a posturated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). In the present report, effects of system pressure on reflooding phenomena shall be discussed based on the data of Tests S1-SH2, S1-01 and S1-02 which are the parameteris tests for system pressure effects belonging to the SCTF Core-I forced flooding test series. Major items discussed in this report are (1) hydrodynamic behavior in the system, (2) core thermal behavior, (3) core heat transfer and (4) two-dimensional hydrodynamic behavior in the pressure vessel including the core. (author)

A study of entrainment at a break and in the core during small break loss-of-coolant accidents in PWRs

International Nuclear Information System (INIS)

Yonomoto, Taisuke

1996-05-01

Objectives of the present study are to obtain a better understanding of entrainment at a break and in the core during small break loss-of-coolant-accidents (SBLOCAs) in PWRs, and to develop a means for the best evaluation of the phenomena. For the study of entrainment at a break, a theoretical model was developed, which was assessed by comparisons with several experimental data bases. By modifying a LOCA analysis code using the present model, experimental results obtained from SBLOCA experiments at a PWR large-scale simulator were reproduced very well. For the study of entrainment in the core, reflooding experiments were conducted at high pressure, from which the onset conditions were obtained. It was confirmed that the cooling behavior for a dry-out core is very simple under typical high pressure reflooding conditions for PWRs, because liquid entrainment does not occur in the core. (author)

SCDAP/RELAP5 code development and assessment

International Nuclear Information System (INIS)

Allison, C.M.; Hohorst, J.K.

1996-01-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The current version of the code is SCDAP/RELAP5/MOD3.1e. Although MOD3.1e contains a number of significant improvements since the initial version of MOD3.1 was released, new models to treat the behavior of the fuel and cladding during reflood have had the most dramatic impact on the code's calculations. This paper provides a brief description of the new reflood models, presents highlights of the assessment of the current version of MOD3.1, and discusses future SCDAP/RELAP5/MOD3.2 model development activities

CFD analysis of blockage length on a partially blocked fuel rod

Energy Technology Data Exchange (ETDEWEB)

Scuro, Nikolas Lymberis; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Angelo, Gabriel [Centro Universitário FEI (UNIFEI), São Paulo, SP (Brazil). Dept. de Engenharia Mecânica; Angelo, Edvaldo, E-mail: nikolas.scuro@gmail.com, E-mail: delvonei@ipen.br, E-mail: gangelo@fei.edu.br, E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, São Paulo, SP (Brazil). Escola da Engenharia. Grupo de Simulação Numérica

2017-07-01

In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod. (author)

Project report - an overview of the project and experiences with project management

DEFF Research Database (Denmark)

Jørgensen, Michael Søgaard; Mikkelsen, Bent Egberg

1996-01-01

A collection of the project planning and the experiences with project management from the Catering 2000 project.As appendieces articles etc. from journals, newspapers etc. about the project.......A collection of the project planning and the experiences with project management from the Catering 2000 project.As appendieces articles etc. from journals, newspapers etc. about the project....

Reseñas. Books Review

Directory of Open Access Journals (Sweden)

Díaz, Liliana

2014-10-01

Full Text Available TORRE, Claudia, El otro desierto de la nación argentina. Antología de narrativa expedicionaria, Bernal, Argentina, Editorial Universidad de Quilmes, 2011, 387 pp. MAURO, Diego y LICHTMAJER, Leandro (compiladores, Los costos de la política. Del Centenario al primer peronismo, Imago Mundi, Buenos Aires, 2014, 130 pp. DE MARCO; Miguel Ángel (h. Ciudad Puerto, Universidad y Desarrollo Regional, Rosario, 1919-1968. CEHDRE, Rosario, 2013, 542 pp. GUARDIA, Sara Beatriz. Primer Congreso Internacional. Las mujeres en los procesos de Independencia de América Latina. Lima, Perú. Sara Beatriz Guardia Edición, CEMHAL, UNESCO, USMP, 2014. 495 pp

Implementation of Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Health Authority by the Agency for Toxic Substances and Disease Registry

International Nuclear Information System (INIS)

Siegel, M.R.

1990-01-01

The Superfund Amendments and Reauthorization Act (SARA) of 1986 greatly expanded the health authority of the Comprehensive Environmental Response, Compensation, and Liability Act. One of the federal agencies most affected by SARA is the Agency for Toxic Substances and Disease Registry (ATSDR) of the U.S. Public Health Service. Among other responsibilities, ATSDR was mandated to conduct health assessments within strict time frames for each site on or proposed for the U.S. Environmental Protection Agency's National Priorities List. The author will review ATSDR's efforts to address this new statutory mandate, especially for federal facilities, and will focus on different conceptual frameworks for implementing the health assessment program

One project`s waste is another project`s resource

Energy Technology Data Exchange (ETDEWEB)

Short, J.

1997-02-01

The author describes the efforts being made toward pollution prevention within the DOE complex, as a way to reduce overall project costs, in addition to decreasing the amount of waste to be handled. Pollution prevention is a concept which is trying to be ingrained into project planning. Part of the program involves the concept that ultimately the responsibility for waste comes back to the generator. Parts of the program involve efforts to reuse materials and equipment on new projects, to recycle wastes to generate offsetting revenue, and to increase awareness, accountability and incentives so as to stimulate action on this plan. Summaries of examples are presented in tables.

RELAP5/MOD2 implementation on various mainframes including the IBM and SX-2 supercomputer

International Nuclear Information System (INIS)

DeForest, D.L.; Hassan, Y.A.

1987-01-01

The RELAP5/MOD2 (cycle 36.04) code is a one-dimensional, two-fluid, nonequilibrium, nonhomogeneous transient analysis code designed to simulate operational and accident scenarios in pressurized water reactors (PWRs). System models are solved using a semi-implicit finite difference method. The code was developed at EG and G in Idaho Falls under sponsorship of the US Nuclear Regulatory Commission (NRC). The major enhancement from RELAP5/MOD1 is the use of a six-equation, two-fluid nonequilibrium and nonhomogeneous model. Other improvements include the addition of a noncondensible gas component and the revision and addition of drag formulation, wall friction, and wall heat transfer. Several test cases were run to benchmark the IBM and SX-2 installations against the CDC computer and the CRAY-2 and CRAY/XMP. These included the Edward's pipe blow-down and two separate reflood cases developed to simulate the FLECHT-SEASET reflood test 31504 and a postcritical heat flux (CHF) test performed at Lehigh University

PREMIUM - Benchmark on the quantification of the uncertainty of the physical models in the system thermal-hydraulic codes

International Nuclear Information System (INIS)

Skorek, Tomasz; Crecy, Agnes de

2013-01-01

PREMIUM (Post BEMUSE Reflood Models Input Uncertainty Methods) is an activity launched with the aim to push forward the methods of quantification of physical models uncertainties in thermal-hydraulic codes. It is endorsed by OECD/NEA/CSNI/WGAMA. The benchmark PREMIUM is addressed to all who applies uncertainty evaluation methods based on input uncertainties quantification and propagation. The benchmark is based on a selected case of uncertainty analysis application to the simulation of quench front propagation in an experimental test facility. Application to an experiment enables evaluation and confirmation of the quantified probability distribution functions on the basis of experimental data. The scope of the benchmark comprises a review of the existing methods, selection of potentially important uncertain input parameters, preliminary quantification of the ranges and distributions of the identified parameters, evaluation of the probability density function using experimental results of tests performed on FEBA test facility and confirmation/validation of the performed quantification on the basis of blind calculation of Reflood 2-D PERICLES experiment. (authors)

TRAC development at General Electric

International Nuclear Information System (INIS)

Andersen, J.G.M.; Shaug, J.C.; Shiralkar, B.S.

1987-01-01

TRAC is a computer code for transient analysis of light water reactors. The BWR version of TRAC has been developed as a result of a close cooperation between General Electric Company and Idaho National Engineering Laboratory. Up through 1985 the development work at General Electric was jointly funded by General Electric, the Nuclear Regulatory Commission and Electric Power Research Institute under the Refill-Reflood and FIST programs. At INEL (which has the main responsibility for the NRC version of TRAC-BWR) this work has led to the development of TRACBD1 and TRACBF1, while at GE, TRACB04 was the final product of the Refill-Reflood and FIST programs. TRAC development has continued at General Electric after the completion of these programs with the evolution of the TRACG code. The purpose of the paper is to describe this work. The TRAC development at General Electric can be divided into two main categories: extended benchmark capability and improved user convenience

Computer codes for the study of the loss of coolant accident of PWR reactors

International Nuclear Information System (INIS)

Gomolinski, M.; Menessier, D.; Tellier, N.

1975-01-01

The CEA has undertaken a large programme to study the consequence on the core of the LOCA of a PWR. In the programme, simultaneously carried out experiments and the development of the calculations means are described. Several experiments such as OMEGA, ERSEC and PHEBUS tests, which provide data to check the computer codes are outlined briefly in the paper. For analysis of the LOCA of a PWR, a series of computer codes, which are at present in use or under development, are linked with each other. The codes are DANAIDES for blowdown, CERES for refill and reflood, THETA-1B and FLIRA for heat up calculation during the blow-down and the reflooding period respectively. FLIRA-PASTEL, a combination of FLIRA and PASTEL which calculate the stress and deformations of material using the finite element method, will be used in place of FLIRA. The basic models and flowcharts of the above codes are described in the paper

Project Success in IT Project Management

OpenAIRE

Siddiqui, Farhan Ahmed

2010-01-01

The rate of failed and challenged Information Technology (IT) projects is too high according to the CHAOS Studies by the Standish Group and the literature on project management (Standish Group, 2008). The CHAOS Studies define project success as meeting the triple constraints of scope, time, and cost. The criteria for project success need to be agreed by all parties before the start of the project and constantly reviewed as the project progresses. Assessing critical success factors is another ...

Vague project start makes project success of outsourced software development projects uncertain

OpenAIRE

Savolainen, Paula

2010-01-01

peer-reviewed A definition of a project success includes at least three criteria: 1) meeting planning goals, 2) customer benefits, and 3) supplier benefits. This study aims to point out the importance of the definition of the project start, the project start date, and what work should be included in the project effort in order to ensure the supplier's benefits. The ambiguity of the project start risks the profitability of the project and therefore makes project success at least from suppli...

Project management: a case of fixed price IS/IT projects. Analysis of projects by project scopes

Directory of Open Access Journals (Sweden)

Miroslav Kral

2012-10-01

Full Text Available The paper provides an overview of major issues of IS / IT projects. Attention will be focused on projects that are implemented under a contract for a specified amount of work and fixed price. The main purpose of the paper is to analyse the project parameters in terms of the types of projects, and to confirm, or refuse, a hypothesis related to this. There is some evidence from the portfolio of projects that have been implemented by the international companies providing IT services. Regarding the localisation, CEE region was selected for our research. The outputs of the paper should be a contribution to managing IS/IT projects in IT service delivery organizations and for the support of innovative thinking about project management generally.

Spain's marketing sector seeing more changes

International Nuclear Information System (INIS)

Anon.

1992-01-01

This paper reports that Spain's petroleum marketing sector continues to restructure. Partly state owned Repsol SA and Royal Dutch/Shell Group are discussing supplying each other's retail outlets in the UK and Spain. And Portugal's state owned Petroleos de Portugal (Petrogal), seeking to sharply expand retail operations in Spain, complains of government interference with foreign investment in Spanish marketing. Meantime, Conoco Inc. Has agreed with Saras SpA Raffinerie Sarde, Milan, to set up a network of service stations in northern Spain and Portugal at a cost of 100 billion pesetas (%972 million). The two are considering building an oil terminal at the port city of Gijon in Asturias, Spain, and the Exxon Corp., Total, and Shell are interested in participating in the project

Managing project complexity : A study into adapting early project phases to improve project performance in large engineering projects

NARCIS (Netherlands)

Bosch-Rekveldt, M.G.C.

2011-01-01

Engineering projects become increasingly more complex and project complexity is assumed to be one of the causes for projects being delivered late and over budget. However, what this project complexity actually comprised of was unclear. To improve the overall project performance, this study focuses

A survey of formal methods for determining functional joint axes.

Science.gov (United States)

Ehrig, Rainald M; Taylor, William R; Duda, Georg N; Heller, Markus O

2007-01-01

Axes of rotation e.g. at the knee, are often generated from clinical gait analysis data to be used in the assessment of kinematic abnormalities, the diagnosis of disease, or the ongoing monitoring of a patient's condition. They are additionally used in musculoskeletal models to aid in the description of joint and segment kinematics for patient specific analyses. Currently available methods to describe joint axes from segment marker positions share the problem that when one segment is transformed into the coordinate system of another, artefacts associated with motion of the markers relative to the bone can become magnified. In an attempt to address this problem, a symmetrical axis of rotation approach (SARA) is presented here to determine a unique axis of rotation that can consider the movement of two dynamic body segments simultaneously, and then compared its performance in a survey against a number of previously proposed techniques. Using a generated virtual joint, with superimposed marker error conditions to represent skin movement artefacts, fitting methods (geometric axis fit, cylinder axis fit, algebraic axis fit) and transformation techniques (axis transformation technique, mean helical axis, Schwartz approach) were classified and compared with the SARA. Nearly all approaches were able to estimate the axis of rotation to within an RMS error of 0.1cm at large ranges of motion (90 degrees ). Although the geometric axis fit produced the least RMS error of approximately 1.2 cm at lower ranges of motion (5 degrees ) with a stationary axis, the SARA and Axis Transformation Technique outperformed all other approaches under the most demanding marker artefact conditions for all ranges of motion. The cylinder and algebraic axis fit approaches were unable to compute competitive AoR estimates. Whilst these initial results using the SARA are promising and are fast enough to be determined "on-line", the technique must now be proven in a clinical environment.

Spent Nuclear Fuel project, project management plan

International Nuclear Information System (INIS)

Fuquay, B.J.

1995-01-01

The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

Managing projects using a project management approach

Directory of Open Access Journals (Sweden)

Marko D. Andrejić

2011-04-01

Full Text Available Modern management theory treats all complex tasks and duties like projects and make these projects possible to be managed by a particular organizational-management concept in order to achieve a goal effectively. A large number of jobs and tasks performed in the system of defense or for defense purposes have the characteristics of projects. Project management is both a skill and a science of monitoring human, material, financial, energy and other resources to achieve required objectives within the given limits: deadlines, time, budget, possibility of realization and the satisfaction of the interests of all project participants. Project management is a traditional area of applied (or functional management focused on managing complex and uncertain situations with defined goals. Introduction In conditions of rapid change and high uncertainty, only adaptive organizations survive, i. e. those that are able not only to react quickly to changes but also to proactively take advantage of changes. Development of project management The biggest influence on the development of the area had complex jobs within the engineering profession. In parallel with the traditional approach new approaches began to develop, while the traditional one still remained in use. Contrary to the traditional engineering approach, a dynamic model first developed in order to respond to demands for greater control of costs. Project management Project management is a skill and knowledge of human and material resources to achieve set objectives within prescribed limits: deadlines, time, budget, possibility of realization, and the satisfaction of all participants in the project. In order to realize a project effectively, it is necessary to manage it rationally. Planning and project management A project plan is a document that allows all team members insight on where to go, when to start and when to arrive, what is necessary to be done in order to achieve the project objectives and what

lessons learned from the QUENCH program at FZK

International Nuclear Information System (INIS)

Steinbrueck, M.; Grosse, M.; Sepold, L.; Stuckert, J.

2011-01-01

The paper gives an overview on the main outcome of the QUENCH program at FZK, including complementary bundle experiments and separate-effects tests. The major objective of the program is to deliver experimental and analytical data to support development and validation of quench and quench-related models as used in code systems. So far, 15 integral bundle QUENCH experiments with 21-31 electrically heated fuel rod simulators of 2.5 m length have been conducted. The following parameters and their influence on bundle degradation and reflood have been investigated: degree of pre-oxidation, temperature at initiation of reflood, flooding rate, influence of neutron absorber materials (B 4 C, AgInCd), air ingress, and the influence of the type of cladding alloy. In six tests reflood of the bundle caused a temporary temperature excursion connected with the release of a significant amount of hydrogen, typically 2 orders of magnitude greater than in those tests with 'successful' quenching in which cool-down was immediately achieved. Comprehensive formation, relocation, and oxidation of melt were observed in all tests with escalation. The temperature boundary between rapid cooldown and temperature escalation was typically 2100-2200 K in the 'normal' quench tests, i.e. tests without absorber and/or steam starvation. Tests with absorber and/or steam starvation were found to lead to temperature escalations at lower temperatures. All phenomena occurring in the bundle tests have been additionally investigated in parametric and more systematic separate-effects tests. Oxidation kinetics of various cladding alloys, including advanced ones, have been determined over a wide temperature range (873-1773 K) in different atmospheres (steam, oxygen, air, and their mixtures). Hydrogen absorption by different zirconium alloys was investigated in detail, recently also using neutron radiography as non-destructive method for determination of hydrogen distribution in claddings

Effects of Parallel Channel Interactions, Steam Flow, Liquid Subcool ...

African Journals Online (AJOL)

Tests were performed to examine the effects of parallel channel interactions, steam flow, liquid subcool and channel heat addition on the delivery of liquid from the upper plenum into the channels and lower plenum of Boiling Water Nuclear Power Reactors during reflood transients. Early liquid delivery into the channels, ...

Project management of life-science research projects: project characteristics, challenges and training needs.

Science.gov (United States)

Beukers, Margot W

2011-02-01

Thirty-four project managers of life-science research projects were interviewed to investigate the characteristics of their projects, the challenges they faced and their training requirements. A set of ten discriminating parameters were identified based on four project categories: contract research, development, discovery and call-based projects--projects set up to address research questions defined in a call for proposals. The major challenges these project managers are faced with relate to project members, leadership without authority and a lack of commitment from the respective organization. Two-thirds of the project managers indicated that they would be interested in receiving additional training, mostly on people-oriented, soft skills. The training programs that are currently on offer, however, do not meet their needs. Copyright © 2010 Elsevier Ltd. All rights reserved.

Source term calculations - Ringhals 2 PWR

International Nuclear Information System (INIS)

Johansson, L.L.

1998-02-01

This project was performed within the fifth and final phase of sub-project RAK-2.1 of the Nordic Co-operative Reactor Safety Program, NKS.RAK-2.1 has also included studies of reflooding of degraded core, recriticality and late phase melt progression. Earlier source term calculations for Swedish nuclear power plants are based on the integral code MAAP. A need was recognised to compare these calculations with calculations done with mechanistic codes. In the present work SCDAP/RELAP5 and CONTAIN were used. Only limited results could be obtained within the frame of RAK-2.1, since many problems were encountered using the SCDAP/RELAP5 code. The main obstacle was the extremely long execution times of the MOD3.1 version, but also some dubious fission product calculations. However, some interesting results were obtained for the studied sequence, a total loss of AC power. The report describes the modelling approach for SCDAP/RELAP5 and CONTAIN, and discusses results for the transient including the event of a surge line creep rupture. The study will probably be completed later, providing that an improved SCDAP/RELAP5 code version becomes available. (au) becomes available. (au)

Source term calculations - Ringhals 2 PWR. Interim report

Energy Technology Data Exchange (ETDEWEB)

Johansson, Lise-Lotte

1998-03-01

This project was performed within the fifth and final phase of sub-project RAK-2.1 of the Nordic Co-operative Reactor Safety Program, NKS. RAK-2.1 has also included studies of reflooding of degraded core, recriticality and late phase melt progression. Earlier source term calculations for Swedish nuclear power plants are based on the integral code MAAP. A need was recognised to compare these calculations with calculations done with mechanistic codes. In the present work SCDAP/RELAP5 and CONTAIN were used. Only limited results could be obtained within the frame of RAK-2.1, since many problems were encountered using the SCDAP/RELAP5 code. The main obstacle was the extremely long execution times of the MOD3.1 version, but also some dubious fission product calculations. However, some interesting results were obtained for the studied sequence, a total loss of AC power. The report describes the modelling approach for SCDAP/RELAP5 and CONTAIN, and discusses results for the transient including the event of a surge line creep rupture. The study will probably be completed later, providing that an improved SCDAP/RELAP5 code version becomes available 8 refs, 16 figs, 5 tabs

Impact of Project Leadership Facets on Project Outcome

Directory of Open Access Journals (Sweden)

Arslan Ayub

2015-08-01

Full Text Available The study analyzes the role of project leadership facets on effective project outcome. Numerous such initiatives have already been taken on project outcome/performance in the context of apposite leadership styles or project management. However, the current study is unique in the milieu of project outcome that it introduces a new leadership approach, which throws light on the significance of variant leadership facets on project outcome. The study uses explanatory approach; primary data is collected from project management professionals working in different project organizations. The study uses structural equation model (SEM technique to test the hypothesis. The study found a positive relationship between project leadership facets and project outcome.

Dynamics of quench front during emergency cooling of PWR core after LOCA accident

International Nuclear Information System (INIS)

Osmachkin, V.S.

2000-01-01

A review of some analytical results for assessment of quenches velocity is presented. Attention is paid to the influence on front velocity due to the peculiarities of fuel, gas gap and fuel pellets as well due to the decay heat and renewed heat-up coming from the cladding oxidation during reflooding. (author)

Project management in practice : Evaluating a case project through project management theories

OpenAIRE

Uusitalo, Jenni

2013-01-01

The purpose of this thesis was to evaluate a case project and to study whether it was carried out in a correct manner; meaning that did the case project follow the project management models. In addition, part of the study was to determine what could have been improved in the management of the case project. The case project was about creating and launching a communication channel based on a social media service, on a blog platform called Tumblr, for Team Finland in Spain network. The network p...

the capacity of megasphaera elsdenii strain ncimb 41125 to control ...

African Journals Online (AJOL)

David

SARA include grain treatment to reduce starch degradation, feed additives such as ... Keywords: Megasphaera elsdenii, ruminal acidosis, dairy cattle, beef cattle, sheep, ...... Mastitis. Retained foetal membranes. Metritis/endometritis. Ketosis.

From project management to project leadership

NARCIS (Netherlands)

Braun, F.; Avital, M.

2010-01-01

It is virtually a truism that good leadership practices can help project managers with attaining the desired project outcome. However, a better understanding of which leadership practices enable project managers to be more effective warrants further investigation. Subsequently, in this study, we

Ace Project as a Project Management Tool

Science.gov (United States)

Cline, Melinda; Guynes, Carl S.; Simard, Karine

2010-01-01

The primary challenge of project management is to achieve the project goals and objectives while adhering to project constraints--usually scope, quality, time and budget. The secondary challenge is to optimize the allocation and integration of resources necessary to meet pre-defined objectives. Project management software provides an active…

76 FR 11494 - List of Recipients of Indian Health Scholarships Under the Indian Health Scholarship Program

Science.gov (United States)

2011-03-02

... Collins, Sara Jane, University of Oklahoma, Cherokee Nation, Oklahoma Cook, David D., Rocky Vista... University, Oglala Sioux Tribe of the Pine Ridge Reservation, South Dakota Jim, Leroy, The Fielding Institute...

Projective-anticipating, projective and projective-lag synchronization of chaotic systems with time-varying delays

International Nuclear Information System (INIS)

Feng Cunfang; Guan Wei; Wang Yinghai

2013-01-01

We investigate different types of projective (projective-anticipating, projective and projective-lag) synchronization in unidirectionally nonlinearly coupled time-delayed chaotic systems with variable time delays. Based on the Krasovskii–Lyapunov approach, we find both the existence and sufficient stability conditions, using a general class of time-delayed chaotic systems related to optical bistable or hybrid optical bistable devices. Our method has the advantage that it requires only one nonlinearly coupled term to achieve different types of projective synchronization in time-delayed chaotic systems with variable time delays. Compared with other existing works, our result provides an easy way to achieve projective-anticipating, projective and projective-lag synchronization. Numerical simulations of the Ikeda system are given to demonstrate the validity of the proposed method. (paper)

Spectral characterization of crude oil using fluorescence (synchronous and time-resolved) and NIR (Near Infrared Spectroscopy); Caracterizacao espectral do petroleo utilizando fluorescencia (sincronizada e resolvida no tempo) e NIR (Near Infrared Spectroscopy)

Energy Technology Data Exchange (ETDEWEB)

Falla Sotelo, F.; Araujo Pantoja, P.; Lopez-Gejo, J.; Le Roux, G.A.C.; Nascimento, C.A.O. [Universidade de Sao Paulo (USP), SP (Brazil). Dept. de Engenharia Quimica. Lab. de Simulacao e Controle de Processos; Quina, F.H. [Universidade de Sao Paulo (USP), SP (Brazil). Inst. de Quimica. Centro de Capacitacao e Pesquisa em Meio Ambiente (CEPEMA)

2008-07-01

The objective of the present work is to evaluate the performance of two spectroscopic techniques employed in the crude oil characterization: NIR spectroscopy and fluorescence spectroscopy (Synchronous fluorescence - SF and Time Resolved Fluorescence - TRF) for the development of correlation models between spectral profiles of crude oil samples and both physical properties (viscosity and API density) and physico-chemical properties (SARA analysis: Saturated, Aromatic, Resins and Asphaltenes). The better results for viscosity and density were obtained using NIR whose prediction capacity was good (1.5 cP and 0.5 deg API, respectively). For SARA analysis, fluorescence spectroscopy revealed its potential in the model calibration showing good results (R2 coefficients greater than 0.85). TRF spectroscopy had better performance than SF spectroscopy. (author)

Application of flow-controllable accumulator and performance analysis in Korean Nuclear Power Plants

International Nuclear Information System (INIS)

Jung, Byung-Ryul; Lee, Un-Chul

1997-01-01

The Korean Yonggwang Nuclear Power Plants 3 ampersand 4(YGN 3 ampersand 4) are the two-loop pressurized water reactor (PWR) nuclear steam supply systems, rated at 2,815 MW(thermal). They incorporate the safety injection system (SIS) consisting of the two high pressure (HPSI) pumps, two low pressure safety injection (LPSI) pumps, and four accumulators. The SIS is two headered arrangements, each to four cold legs injection (CLI) type which provides cooling to the core in the highly unlikely event of a loss-of-coolant accident (LOCA). In the current SIS, the LPSI pumps automatically start during a LOCA, and also provide the residual heat removal capability during the shutdown cooling. This paper presents the feasibility of the removal of the LPSI from the existing SIS with minor system changes, including the increase up to four in the HPSI pumps, direct vessel injection(DVI), and the flow-controllable accumulators. A double-ended rupture of one of the four cold legs in the YGN 3 ampersand 4 was simulated using RELAP5/MOD3.1 to determine the feasibility of the application of this new SIS design to the current nuclear power plants. As a result, the calculated reflooding peak cladding surface temperature(PCT) was comparable to that of original base calculation, and the downcomer and the core collapsed liquid level during reflooding were also comparable to those in the current safety system design. This large break, cold-leg LOCA analysis addresses the reflooding capability without credit for a LPSI pump system and the applicability of the new flow-controllable accumulator. Also this analysis confirms that the combination of new flow-controllable accumulators, DVI and the increased HPSI pumps maintain the peak cladding temperature below the prescribed limits. 14 refs., 4 figs., 3 tabs

Project Success in Agile Development Software Projects

Science.gov (United States)

Farlik, John T.

2016-01-01

Project success has multiple definitions in the scholarly literature. Research has shown that some scholars and practitioners define project success as the completion of a project within schedule and within budget. Others consider a successful project as one in which the customer is satisfied with the product. This quantitative study was conducted…

An IS Project Management Course Project

Science.gov (United States)

Frank, Ronald L.

2010-01-01

Information Systems curricula should provide project management (PM) theory, current practice, and hands-on experience. The schedule usually does not allow time in Analysis and Design courses for development oriented project management instruction other than a short introduction. Similarly, networking courses usually don't put project management…

Evaluation of Project Achievements in VOMARE -project

OpenAIRE

Kokkarinen, Eeva

2011-01-01

The purpose of the thesis is to study the achievements of VOMARE –project from the Finnish Lifeboat Institutions perspective. The organisation is a roof organisation for voluntary maritime rescue operation in Finland. The Finnish Lifeboat Institution is a lead partner in VOMARE –project which is EU funded project and the aim of the project is to start voluntary rescue operations in Estonia. The theoretical part of the work is divided into two main categories; project management and planni...

Leading global projects for professional and accidental project leaders

CERN Document Server

Moran, Robert T

2008-01-01

This book is a must-read for anyone responsible for projects and initiatives that span functional and geographical divides. Authors Moran and Youngdahl bring extensive experience and learning from industry practice to present a clear and straightforward treatment of the leadership skills and knowledge required to lead projects that are global in nature. They have written the first book of its kind to address the three essential skills of global project leaders - strategic project management, project leadership, and cross-cultural leadership. The authors argue that global project leadership is an essential skill in our project-based world and that we are all either intentional or accidental project leaders. Intentional project leaders pursue formal project management education and even certification whereas accidental project leaders find themselves leading global projects and initiatives as a result of a special assignment or promotion. Moran and Youndahl have found that the vast majority of global projects ...

Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

1993-03-01

As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

Experimental investigation of the coolability of blocked hexagonal bundles

Energy Technology Data Exchange (ETDEWEB)

Hózer, Zoltán, E-mail: zoltan.hozer@energia.mta.hu; Nagy, Imre; Kunstár, Mihály; Szabó, Péter; Vér, Nóra; Farkas, Róbert; Trosztel, István; Vimi, András

2017-06-15

Highlights: • Experiments were performed with electrically heated hexagonal fuel bundles. • Coolability of ballooned VVER-440 type bundle was confirmed up to high blockage rate. • Pellet relocation effect causes delay in the cool-down of the bundle. • The bypass line does not prevent the reflood of ballooned fuel rods. - Abstract: The CODEX-COOL experimental series was carried out in order to evaluate the effect of ballooning and pellet relocation in hexagonal bundles on the coolability of fuel rods after a LOCA event. The effects of blockage geometry, coolant flowrate, initial temperature and axial profile were investigated. The experimental results confirmed that a VVER bundle up to 80% blockage rate remains coolable after a LOCA event under design basis conditions. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front.

Evaluation of the RELAP4/MOD6 thermal-hydraulic code

International Nuclear Information System (INIS)

Haigh, W.S.; Margolis, S.G.; Rice, R.E.

1978-01-01

The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA

Prioritising Project Scope Definition Elements in Public Building Projects

Directory of Open Access Journals (Sweden)

Mohammed K Fageha

2014-09-01

Full Text Available A complete definition of the scope of a project upfront during early stages ensures smooth and successful implementation during the project execution. This research identifies and prioritises project scope definition elements for public buildings in Saudi Arabia. Elements that could significantly contribute to complete project scope definition package at pre-project planning stage are identified and their interrelationship determined and prioritised. Using the Project Definition Rating Index (PDRI as a basis, the study uses analytical network process (ANP technique based on data obtained from project managers who have been involved in public sector projects in Saudi Arabia. Data collection and analysis was conducted in three steps. The first step involved identification of scope definition elements while the second involved an investigation into interrelationships among the elements. In the third step, ANP was used to determine the weight of the elements’ importance in terms of contribution to project scope definition completeness. Finally, Pareto analysis was used to prioritise and assess the distribution pattern of the elements. The outcome from this research is the prioritisation of project scope definition elements for public building projects in Saudi Arabia. The prioritised list developed indicates the importance of project scope definition elements. It should help project management teams identify elements to consider when evaluating project scope definition for completeness at the pre-project planning stage. Keywords: Project scope definition, pre-project planning, prioritising, public building projects, Saudi Arabia, Analytical Network Process (ANP

Bacteriophage-antibiotic synergism to control planktonic and biofilm ...

African Journals Online (AJOL)

Bacteriophage-antibiotic synergism to control planktonic and biofilm producing clinical isolates of Pseudomonas aeruginosa. Amina Amal Mahmoud Nouraldin, Manal Mohammad Baddour, Reem Abdel Hameed Harfoush, Sara AbdelAziz Mohamed Essa ...

The application of project management in construction projects ...

African Journals Online (AJOL)

Project management is critical for successful project development. A crucial responsibility of the project manager is ensuring that the client is certified and the scope of work is of high quality, within the agreed budget and time frame. In some way, project financing is completed from the time of project conception. Indeed ...

Perspectives on projects, project success and team work

OpenAIRE

Thompson, Karen

2015-01-01

This paper brings together perspectives on projects, project success and team work as a background to two graphical tools for considering project success and individual capabilities for working in a project team.

Project Management Yinyang: Coupling project success and client satisfaction

Directory of Open Access Journals (Sweden)

Greg Stewart Usher

2017-06-01

Full Text Available Our research applies paradox theory to a project management construct to help project management researchers and practitioners understand the tensions that can exist between project success and client satisfaction. Our research highlights that although project success and client satisfaction are both present within a project management construct, they also belong to different functional systems. Project success and client satisfaction have different systemic-discourses and use different language games to convey information. These distinctions can create latent and sometimes salient tensions within the project management construct that project managers must understand, embrace, and work with. We have used a Grounded Theory (GT methodology to explore the lived experience of project managers, and from this have identified a phenomenon which we have termed project management yinyang. Project management yinyang is the state that exists when both project success and Client satisfaction are tightly coupled within the project management construct. Project management yinyang highlights that these two phenomena cannot be viewed as separate elements because the ‘seed’ of each exists within the other. And to truly achieve one, you must also achieve the other. Our findings indicate that in order to create project management yinyang the project manager must embrace a paradoxical yet holistic philosophy. They must understand the complementarity, interdependency, and structural coupling that exists between the positivist and interpretivist paradigms within the project management construct. They must understand how satisfaction (Yin and success (Yang are created through focus. Furthermore, they must understand how project management yinyang is separate from, but borne from, the convergence of the other two elements.

Quality Assurance Project Plan for Citizen Science Projects

Science.gov (United States)

The Quality Assurance Project Plan is necessary for every project that collects or uses environmental data. It documents the project planning process and serves as a blueprint for how your project will run.

Pop / Tõnis Kahu

Index Scriptorium Estoniae

Kahu, Tõnis, 1962-

2008-01-01

Heliplaatidest: Get Cape. Wear Cape. Fly "Searching For The Hows And Whys", The Black Crowes "Warpaint", Jonny Greenwood "There Will Be Blood", Tegan & Sara "The Con", Teräsbetoni "Myrskyntuoja", REM "Accelerate"

Hella hingega karmid naised / Krista Kaer

Index Scriptorium Estoniae

Kaer, Krista

1996-01-01

Tänapäeva kriminaalromaani inglise ja ameerika autoritest (Dame Phyllis Dorothy James, Joan Smith, Lynda La Plante (inglise); Sara Paretsky, Sue Grafton, Patricia Cornwell (ameerika) ja nende poolt loodud naisdetektiivi kujudest

Policy challenges in modern health care

National Research Council Canada - National Science Library

Mechanic, David

2005-01-01

... for the Obesity Epidemic KENNETH E. WARNER 99 8 Patterns and Causes of Disparities in Health DAVID R. WILLIAMS 115 9 Addressing Racial Inequality in Health Care SARA ROSENBAUM AND JOEL TEITELBAU...

Project Management

DEFF Research Database (Denmark)

Pilkington, Alan; Chai, Kah-Hin; Le, Yang

2015-01-01

This paper identifies the true coverage of PM theory through a bibliometric analysis of the International Journal of Project Management from 1996-2012. We identify six persistent research themes: project time management, project risk management, programme management, large-scale project management......, project success/failure and practitioner development. These differ from those presented in review and editorial articles in the literature. In addition, topics missing from the PM BOK: knowledge management project-based organization and project portfolio management have become more popular topics...

Project financing

International Nuclear Information System (INIS)

Cowan, A.

1998-01-01

Project financing was defined ('where a lender to a specific project has recourse only to the cash flow and assets of that project for repayment and security respectively') and its attributes were described. Project financing was said to be particularly well suited to power, pipeline, mining, telecommunications, petro-chemicals, road construction, and oil and gas projects, i.e. large infrastructure projects that are difficult to fund on-balance sheet, where the risk profile of a project does not fit the corporation's risk appetite, or where higher leverage is required. Sources of project financing were identified. The need to analyze and mitigate risks, and being aware that lenders always take a conservative view and gravitate towards the lowest common denominator, were considered the key to success in obtaining project financing funds. TransAlta Corporation's project financing experiences were used to illustrate the potential of this source of financing

Wisdom for Building the Project Manager/Project Sponsor Relationship: Partnership for Project Success

National Research Council Canada - National Science Library

Patton, Nanette; Shechet, Allan

2007-01-01

.... This article discusses conventional roles and responsibilities of the project sponsor and then discusses strategies a project manager can employ to define boundaries to reduce role confusion and promote partnership to facilitate project success.

The CHPRC Columbia River Protection Project Quality Assurance Project Plan

International Nuclear Information System (INIS)

Fix, N.J.

2008-01-01

Pacific Northwest National Laboratory researchers are working on the CHPRC Columbia River Protection Project (hereafter referred to as the Columbia River Project). This is a follow-on project, funded by CH2M Hill Plateau Remediation Company, LLC (CHPRC), to the Fluor Hanford, Inc. Columbia River Protection Project. The work scope consists of a number of CHPRC funded, related projects that are managed under a master project (project number 55109). All contract releases associated with the Fluor Hanford Columbia River Project (Fluor Hanford, Inc. Contract 27647) and the CHPRC Columbia River Project (Contract 36402) will be collected under this master project. Each project within the master project is authorized by a CHPRC contract release that contains the project-specific statement of work. This Quality Assurance Project Plan provides the quality assurance requirements and processes that will be followed by the Columbia River Project staff

The CHPRC Columbia River Protection Project Quality Assurance Project Plan

Energy Technology Data Exchange (ETDEWEB)

Fix, N. J.

2008-11-30

Pacific Northwest National Laboratory researchers are working on the CHPRC Columbia River Protection Project (hereafter referred to as the Columbia River Project). This is a follow-on project, funded by CH2M Hill Plateau Remediation Company, LLC (CHPRC), to the Fluor Hanford, Inc. Columbia River Protection Project. The work scope consists of a number of CHPRC funded, related projects that are managed under a master project (project number 55109). All contract releases associated with the Fluor Hanford Columbia River Project (Fluor Hanford, Inc. Contract 27647) and the CHPRC Columbia River Project (Contract 36402) will be collected under this master project. Each project within the master project is authorized by a CHPRC contract release that contains the project-specific statement of work. This Quality Assurance Project Plan provides the quality assurance requirements and processes that will be followed by the Columbia River Project staff.

Basics of SCI Rehabilitation

Medline Plus

Full Text Available ... How Peer Counseling Works Julie Gassaway, MS, RN Pediatric Injuries Pediatric Spinal Cord Injury 101 Lawrence Vogel, MD The Basics of Pediatric SCI Rehabilitation Sara Klaas, MSW Transitions for Children ...

Preventing Pressure Sores

Medline Plus

Full Text Available ... How Peer Counseling Works Julie Gassaway, MS, RN Pediatric Injuries Pediatric Spinal Cord Injury 101 Lawrence Vogel, MD The Basics of Pediatric SCI Rehabilitation Sara Klaas, MSW Transitions for Children ...

Sex and Fertility After SCI

Medline Plus

Full Text Available ... How Peer Counseling Works Julie Gassaway, MS, RN Pediatric Injuries Pediatric Spinal Cord Injury 101 Lawrence Vogel, MD The Basics of Pediatric SCI Rehabilitation Sara Klaas, MSW Transitions for Children ...

Spinal Cord Injury 101

Medline Plus

Full Text Available ... How Peer Counseling Works Julie Gassaway, MS, RN Pediatric Injuries Pediatric Spinal Cord Injury 101 Lawrence Vogel, MD The Basics of Pediatric SCI Rehabilitation Sara Klaas, MSW Transitions for Children ...

Journal of Biosciences | Indian Academy of Sciences

Indian Academy of Sciences (India)

Author Affiliations. Kayvan Etebari1 Leila Matindoost2. Department of Sericulture, Faculty of Natural Resources, University of Guilan, Somehe Sara, Iran; Department of Entomology, Faculty of Agriculture, University of Guilan, Rasht, Iran ...

Project structure plan requirements for the deconstruction projects

International Nuclear Information System (INIS)

Petrasch, Peter; Schmitt, Christian; Stapf, Meike

2011-01-01

The deconstruction of nuclear facilities requires due to the particular conditions and the size of the project a special project planning. The authors analyze the possible requirements to be fulfilled by a project structure plan for nuclear facilities, including personnel resources, organization structure, budget questions, operation and project oriented measures, possibility of modifications and supplements. Further topics include controlling and project realization procedures, documentation, third party activities (authorities, consultants, surveyors), logistics and transport, and radiation protection issues. Several questions remain for plants-specific planning, including the integration of the plant personnel, administrative work, project management, economic and financial issues, radioactive waste management issues.

Adapting Project Management Practices to Research-Based Projects

Science.gov (United States)

Bahr, P.; Baker, T.; Corbin, B.; Keith, L.; Loerch, L.; Mullenax, C.; Myers, R.; Rhodes, B.; Skytland, N.

2007-01-01

From dealing with the inherent uncertainties in outcomes of scientific research to the lack of applicability of current NASA Procedural Requirements guidance documentation, research-based projects present challenges that require unique application of classical project management techniques. If additionally challenged by the creation of a new program transitioning from basic to applied research in a technical environment often unfamiliar with the cost and schedule constraints addressed by project management practices, such projects can find themselves struggling throughout their life cycles. Finally, supplying deliverables to a prime vehicle customer, also in the formative stage, adds further complexity to the development and management of research-based projects. The Biomedical Research and Countermeasures Projects Branch at NASA Johnson Space Center encompasses several diverse applied research-based or research-enabling projects within the newly-formed Human Research Program. This presentation will provide a brief overview of the organizational structure and environment in which these projects operate and how the projects coordinate to address and manage technical requirements. We will identify several of the challenges (cost, technical, schedule, and personnel) encountered by projects across the Branch, present case reports of actions taken and techniques implemented to deal with these challenges, and then close the session with an open forum discussion of remaining challenges and potential mitigations.

Organization of BSc and MSc projects in project families

DEFF Research Database (Denmark)

Ottosen, Lisbeth M.; Goltermann, Per; Jensen, Pernille Erland

2014-01-01

This work reports organization of student thesis projects in project families, with the benefit to both teaching and learning. The project organization went from student projects broadly distributed on topics related to different research issues and individual supervision to project families with...

Subacute ruminal acidosis reduces sperm quality in beef bulls.

Science.gov (United States)

Callaghan, M J; McAuliffe, P; Rodgers, R J; Hernandez-Medrano, J; Perry, V E A

2016-08-01

Breeding bulls are commonly fed high-energy diets, which may induce subacute ruminal acidosis (SARA). In this experiment, 8 Santa Gertrudis bulls (age 20 ± 6 mo) were used to evaluate the extent and duration of effects of SARA on semen quality and the associated changes in circulating hormones and metabolites. The bulls were relocated and fed in yards with unrestricted access to hay and daily individual concentrate feeding for 125 d before SARA challenge. Semen was collected and assessed at 14-d intervals before the challenge to ensure acclimatization and the attainment of a stable spermiogram. The challenge treatments consisted of either a single oral dose of oligofructose (OFF; 6.5 g/kg BW) or an equivalent sham dose of water (Control). Locomotion, behavior, respiratory rate, and cardiovascular and gastrointestinal function were intensively monitored during the 24-h challenge period. Rumen fluid samples were retained for VFA, ammonia, and lactate analysis. After the challenge, semen was then collected every third day for a period of 7 wk and then once weekly until 12 wk, with associated blood collection for FSH, testosterone, inhibin, and cortisol assay. Percent normal sperm decreased in bulls dosed with OFF after the challenge period ( < 0.05) and continued to remain lower on completion of the study at 88 d after challenge. There was a corresponding increase in sperm defects commencing from 16 d after challenge. These included proximal cytoplasmic droplets ( < 0.001), distal reflex midpieces ( = 0.01), and vacuole and teratoid heads ( < 0.001). Changes in semen quality after challenge were associated with lower serum testosterone ( < 0.001) and FSH ( < 0.05). Serum cortisol in OFF bulls tended to be greater ( = 0.07) at 7 d after challenge. This study shows that SARA challenge causes a reduction in sperm quality sufficient to preclude bulls from sale as single sire breeding animals 3 mo after the event occurred.

Pneumonia por varicela associada com síndrome da angústia respiratória aguda: relato de dois casos Varicella pneumonia complicated with acute respiratory distress syndrome: two cases report

Directory of Open Access Journals (Sweden)

Marcelo Moreno

2007-03-01

Full Text Available JUSTIFICATIVA E OBJETIVOS: A varicela é uma doença exantemática causada pela infecção primária do vírus varicela zoster (VVZ. A pneumonia pelo VVZ complicada com a síndrome da angústia respiratória aguda (SARA é rara e associa-se a altas taxas de morbimortalidade. O objetivo deste estudo foi apresentar dois casos de pneumonia por varicela que evoluíram com SARA e outras disfunções orgânicas. RELATO DOS CASOS: Paciente de 15 anos, imunocomprometido com a síndrome da imunodeficiência adquirida (SIDA e uma paciente do sexo feminino imunocompetente, foram admitidos na UTI com quadro clínico de varicela, SARA, trombocitopenia e acidose graves. Além disso, disfunção cardiovascular e falência renal ocorreram no primeiro e segundo casos, respectivamente. Foram tratados com aciclovir além de ventilação mecânica protetora. CONCLUSÕES: Os dois casos de pneumonia por varicela, que apresentaram SARA e disfunções de múltiplos órgãos, obtiveram boa evolução clínica.BACKGROUNG AND OBJECTIVES: Varicella is an exantematic disease caused by varicella-zoster virus. Varicella pneumonia complicated with acute respiratory distress syndrome (ARDS is very rare in adults and is associated with high morbimortality. We report two cases of ARDS secondary to varicella-zoster virus pneumonia. CASES REPORT: We report two cases of ARDS and multiple organ dysfunction syndrome (MODS secondary to varicella-zoster virus pneumonia. A 15-year-old man with human immunodeficiency virus (HIV infection and a 29-year-old immunocompetent female were admitted in the ICU with primary varicella infection and pneumonia. Both cases progressed towards ARDS, severe thrombocytopenia and acidosis. In addition cardiovascular and renal failure occurred in the first and second patients, respectively. Treatment consisted of immediate administration of intravenous acyclovir and a lung-protective ventilation strategy. CONCLUSIONS: Both cases of varicella

Sigma Factor SigB Is Crucial to Mediate Staphylococcus aureus Adaptation during Chronic Infections.

Directory of Open Access Journals (Sweden)

Lorena Tuchscherr

2015-04-01

Full Text Available Staphylococcus aureus is a major human pathogen that causes a range of infections from acute invasive to chronic and difficult-to-treat. Infection strategies associated with persisting S. aureus infections are bacterial host cell invasion and the bacterial ability to dynamically change phenotypes from the aggressive wild-type to small colony variants (SCVs, which are adapted for intracellular long-term persistence. The underlying mechanisms of the bacterial switching and adaptation mechanisms appear to be very dynamic, but are largely unknown. Here, we analyzed the role and the crosstalk of the global S. aureus regulators agr, sarA and SigB by generating single, double and triple mutants, and testing them with proteome analysis and in different in vitro and in vivo infection models. We were able to demonstrate that SigB is the crucial factor for adaptation in chronic infections. During acute infection, the bacteria require the simultaneous action of the agr and sarA loci to defend against invading immune cells by causing inflammation and cytotoxicity and to escape from phagosomes in their host cells that enable them to settle an infection at high bacterial density. To persist intracellularly the bacteria subsequently need to silence agr and sarA. Indeed agr and sarA deletion mutants expressed a much lower number of virulence factors and could persist at high numbers intracellularly. SigB plays a crucial function to promote bacterial intracellular persistence. In fact, ΔsigB-mutants did not generate SCVs and were completely cleared by the host cells within a few days. In this study we identified SigB as an essential factor that enables the bacteria to switch from the highly aggressive phenotype that settles an acute infection to a silent SCV-phenotype that allows for long-term intracellular persistence. Consequently, the SigB-operon represents a possible target to develop preventive and therapeutic strategies against chronic and therapy

Sigma Factor SigB Is Crucial to Mediate Staphylococcus aureus Adaptation during Chronic Infections.

Science.gov (United States)

Tuchscherr, Lorena; Bischoff, Markus; Lattar, Santiago M; Noto Llana, Mariangeles; Pförtner, Henrike; Niemann, Silke; Geraci, Jennifer; Van de Vyver, Hélène; Fraunholz, Martin J; Cheung, Ambrose L; Herrmann, Mathias; Völker, Uwe; Sordelli, Daniel O; Peters, Georg; Löffler, Bettina

2015-04-01

Staphylococcus aureus is a major human pathogen that causes a range of infections from acute invasive to chronic and difficult-to-treat. Infection strategies associated with persisting S. aureus infections are bacterial host cell invasion and the bacterial ability to dynamically change phenotypes from the aggressive wild-type to small colony variants (SCVs), which are adapted for intracellular long-term persistence. The underlying mechanisms of the bacterial switching and adaptation mechanisms appear to be very dynamic, but are largely unknown. Here, we analyzed the role and the crosstalk of the global S. aureus regulators agr, sarA and SigB by generating single, double and triple mutants, and testing them with proteome analysis and in different in vitro and in vivo infection models. We were able to demonstrate that SigB is the crucial factor for adaptation in chronic infections. During acute infection, the bacteria require the simultaneous action of the agr and sarA loci to defend against invading immune cells by causing inflammation and cytotoxicity and to escape from phagosomes in their host cells that enable them to settle an infection at high bacterial density. To persist intracellularly the bacteria subsequently need to silence agr and sarA. Indeed agr and sarA deletion mutants expressed a much lower number of virulence factors and could persist at high numbers intracellularly. SigB plays a crucial function to promote bacterial intracellular persistence. In fact, ΔsigB-mutants did not generate SCVs and were completely cleared by the host cells within a few days. In this study we identified SigB as an essential factor that enables the bacteria to switch from the highly aggressive phenotype that settles an acute infection to a silent SCV-phenotype that allows for long-term intracellular persistence. Consequently, the SigB-operon represents a possible target to develop preventive and therapeutic strategies against chronic and therapy-refractory infections.

Project studies

DEFF Research Database (Denmark)

Geraldi, Joana; Söderlund, Jonas

2018-01-01

Project organising is a growing field of scholarly inquiry and management practice. In recent years, two important developments have influenced this field: (1) the study and practice of projects have extended their level of analysis from mainly focussing on individual projects to focussing on micro......, and of the explanations of project practices they could offer. To discuss avenues for future research on projects and project practice, this paper suggests the notion of project studies to better grasp the status of our field. We combine these two sets of ideas to analyse the status and future options for advancing...... project research: (1) levels of analysis; and (2) type of research. Analysing recent developments within project studies, we observe the emergence of what we refer to as type 3 research, which reconciles the need for theoretical development and engagement with practice. Type 3 research suggests pragmatic...

A phenomenological analysis of melt progression in the lower head of a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Seiler, J.M., E-mail: jean-marie.seiler@cea.fr [CEA, DEN, DTN, F-38054 Grenoble (France); Tourniaire, B. [EDF/Septen, Lyon (France)

2014-03-15

Highlights: • We propose a phenomenological description of melt progression into the lower head. • We examine changes in heat loads on the vessel. • Heat loads are more severe than emphasized by the bounding situation assumption. • Both primary circuit and ex-vessel reflooding are necessary for in-vessel retention. • Vessel failure conditions are examined. - Abstract: The analysis of in-vessel corium cooling (IVC) and retention (IVR) involves the description of very complex and transient physical phenomena. To get round this difficulty, “bounding” situations are often emphasized for the demonstration of corium coolability, by vessel flooding and/or by reactor pit flooding. This approach however comes up against its own limitations. More realistic melt progression scenarios are required to provide plausible corium configurations and vessel failure conditions. Work to develop more realistic melt progression scenarios has been done at CEA, in collaboration with EDF. Development has concentrated on the French 1300 MWe PWR, considering both dry scenarios and the possibility of flooding of the RPC (reactor primary circuit) and/or the reactor pit. The models used for this approach have been derived from the analysis of the TMI2 accident and take benefit from the lessons derived from several programs related to pool thermal hydraulics (BALI, COPO, ACOPO, etc.), material interactions (RASPLAV, MASCA), critical heat flux (CHF) on the external surface of the vessel (KAIST, SULTAN, ULPU), etc. Important conclusions of this work are as follows: (a)After the start of corium melting and onset of melt formation in the core at low pressure (∼1 to 5 bars), it seems questionable that RPV (reactor pressure vessel) reflooding alone would be sufficient to achieve corium retention in the vessel; (b)If the vessel is not cooled externally, it may fail due to local heat-up before the whole core fuel inventory is relocated in the lower head; (c)Even if the vessel is

Multiple levels in the organisation of innovation : project organization in single-firm projects and multi-firm projects

NARCIS (Netherlands)

Jaspers, F.P.H.; Ende, van den J.C.M.; Borgh, van der W.; Yin, Jie

2008-01-01

Studies about how the organization of new product (and new service) development projects (NPD) projects influences project performance typically investigate this in Single-firm projects, i.e. projects with high ownership integration. However, NPD projects are often performed by two or more

Investigating Asphaltenes Composition in Crude Oil Samples using ...

African Journals Online (AJOL)

MBI

2015-12-22

Dec 22, 2015 ... composition of asphaltenes by Iatroscan TLC-FID method was compared with the weight% of asphaltenes precipitated. ... SARA in the crude oil samples were determined in this work ..... Fractionation and characterization of.

Mind Over Matter: Methamphetamine

Science.gov (United States)

... Teaching Guide and Series / Methamphetamine Mind Over Matter: Methamphetamine (Meth) Print Order Free Publication in: English Spanish ... paranoia, aggressiveness, and hallucinations. The Brain's Response to Methamphetamine Hi, my name's Sara Bellum. Welcome to my ...

Research Project Evaluation-Learnings from the PATHWAYS Project Experience.

Science.gov (United States)

Galas, Aleksander; Pilat, Aleksandra; Leonardi, Matilde; Tobiasz-Adamczyk, Beata

2018-05-25

Every research project faces challenges regarding how to achieve its goals in a timely and effective manner. The purpose of this paper is to present a project evaluation methodology gathered during the implementation of the Participation to Healthy Workplaces and Inclusive Strategies in the Work Sector (the EU PATHWAYS Project). The PATHWAYS project involved multiple countries and multi-cultural aspects of re/integrating chronically ill patients into labor markets in different countries. This paper describes key project's evaluation issues including: (1) purposes, (2) advisability, (3) tools, (4) implementation, and (5) possible benefits and presents the advantages of a continuous monitoring. Project evaluation tool to assess structure and resources, process, management and communication, achievements, and outcomes. The project used a mixed evaluation approach and included Strengths (S), Weaknesses (W), Opportunities (O), and Threats (SWOT) analysis. A methodology for longitudinal EU projects' evaluation is described. The evaluation process allowed to highlight strengths and weaknesses and highlighted good coordination and communication between project partners as well as some key issues such as: the need for a shared glossary covering areas investigated by the project, problematic issues related to the involvement of stakeholders from outside the project, and issues with timing. Numerical SWOT analysis showed improvement in project performance over time. The proportion of participating project partners in the evaluation varied from 100% to 83.3%. There is a need for the implementation of a structured evaluation process in multidisciplinary projects involving different stakeholders in diverse socio-environmental and political conditions. Based on the PATHWAYS experience, a clear monitoring methodology is suggested as essential in every multidisciplinary research projects.

Project financing

International Nuclear Information System (INIS)

Alvarez, M.U.

1990-01-01

This paper presents the basic concepts and components of the project financing of large industrial facilities. Diagrams of a simple partnership structure and a simple leveraged lease structure are included. Finally, a Hypothetical Project is described with basic issues identified for discussion purposes. The topics of the paper include non-recourse financing, principal advantages and objectives, disadvantages, project financing participants and agreements, feasibility studies, organization of the project company, principal agreements in a project financing, insurance, and an examination of a hypothetical project

Project Notes

Science.gov (United States)

School Science Review, 1978

1978-01-01

Presents sixteen project notes developed by pupils of Chipping Norton School and Bristol Grammar School, in the United Kingdom. These Projects include eight biology A-level projects and eight Chemistry A-level projects. (HM)