Migration of alcator C-Mod computer infrastructure to Linux

International Nuclear Information System (INIS)

Fredian, T.W.; Greenwald, M.; Stillerman, J.A.

2004-01-01

The Alcator C-Mod fusion experiment at MIT in Cambridge, Massachusetts has been operating for twelve years. The data handling for the experiment during most of this period was based on MDSplus running on a cluster of VAX and Alpha computers using the OpenVMS operating system. While the OpenVMS operating system provided a stable reliable platform, the support of the operating system and the software layered on the system has deteriorated in recent years. With the advent of extremely powerful low cost personal computers and the increasing popularity and robustness of the Linux operating system a decision was made to migrate the data handling systems for C-Mod to a collection of PC's running Linux. This paper will describe the new system configuration, the effort involved in the migration from OpenVMS, the results of the first run campaign under the new configuration and the impact the switch may have on the rest of the MDSplus community

Validation of one-dimensional module of MARS 2.1 computer code by comparison with the RELAP5/MOD3.3 developmental assessment results

International Nuclear Information System (INIS)

Lee, Y. J.; Bae, S. W.; Chung, B. D.

2003-02-01

This report records the results of the code validation for the one-dimensional module of the MARS 2.1 thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 code development assessment problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS 2.1 code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The results suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

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)

Assessment of RELAP5/MOD2 and RELAP5/MOD1-EUR codes on the basis of LOBI-MOD2 test results

International Nuclear Information System (INIS)

D'Auria, F.; Mazzini, M.; Oriolo, F.; Galassi, G.M.

1989-10-01

The present report deals with an overview of the application of RELAP5/MOD2 and RELAP5/MOD1-EUR codes to tests performed in the LOBI/MOD2 facility. The work has been carried out in the frame of a contract between Dipartimento di Costruzioni Meccaniche e Nucleari (DCMN) of Pisa University and CEC. The Universities of Roma, Pisa, Bologna and Palermo and the Polytechnic of Torino performed the post-test analysis of the LOBI experiment under the supervision of DCMN. In the report the main outcomes from the analysis of the LOBI experiments are given with the attempt to identify deficiencies in the modelling capabilities of the used codes

Vectorization, parallelization and implementation of nuclear codes [MVP/GMVP, QMDRELP, EQMD, HSABC, CURBAL, STREAM V3.1, TOSCA, EDDYCAL, RELAP5/MOD2/C36-05, RELAP5/MOD3] on the VPP500 computer system. Progress report 1995 fiscal year

International Nuclear Information System (INIS)

Nemoto, Toshiyuki; Watanabe, Hideo; Fujita, Toyozo; Kawai, Wataru; Harada, Hiroo; Gorai, Kazuo; Yamasaki, Kazuhiko; Shoji, Makoto; Fujii, Minoru.

1996-07-01

At Center for Promotion of Computational Science and Engineering, time consuming eight nuclear codes suggested by users have been vectorized, parallelized on the VPP500 computer system. In addition, two nuclear codes used on the VP2600 computer system were implemented on the VPP500 computer system. Neutron and photon transport calculation code MVP/GMVP and relativistic quantum molecular dynamics code QMDRELP have been parallelized. Extended quantum molecular dynamics code EQMD and adiabatic base calculation code HSABC have been parallelized and vectorized. Ballooning turbulence simulation code CURBAL, 3-D non-stationary compressible fluid dynamics code STREAM V3.1, operating plasma analysis code TOSCA and eddy current analysis code EDDYCAL have been vectorized. Reactor safety analysis code RELAP5/MOD2/C36-05 and RELAP5/MOD3 were implemented on the VPP500 computer system. (author)

Vectorization, parallelization and implementation of nuclear codes =MVP/GMVP, QMDRELP, EQMD, HSABC, CURBAL, STREAM V3.1, TOSCA, EDDYCAL, RELAP5/MOD2/C36-05, RELAP5/MOD3= on the VPP500 computer system. Progress report 1995 fiscal year

Energy Technology Data Exchange (ETDEWEB)

Nemoto, Toshiyuki; Watanabe, Hideo; Fujita, Toyozo [Fujitsu Ltd., Tokyo (Japan); Kawai, Wataru; Harada, Hiroo; Gorai, Kazuo; Yamasaki, Kazuhiko; Shoji, Makoto; Fujii, Minoru

1996-06-01

At Center for Promotion of Computational Science and Engineering, time consuming eight nuclear codes suggested by users have been vectorized, parallelized on the VPP500 computer system. In addition, two nuclear codes used on the VP2600 computer system were implemented on the VPP500 computer system. Neutron and photon transport calculation code MVP/GMVP and relativistic quantum molecular dynamics code QMDRELP have been parallelized. Extended quantum molecular dynamics code EQMD and adiabatic base calculation code HSABC have been parallelized and vectorized. Ballooning turbulence simulation code CURBAL, 3-D non-stationary compressible fluid dynamics code STREAM V3.1, operating plasma analysis code TOSCA and eddy current analysis code EDDYCAL have been vectorized. Reactor safety analysis code RELAP5/MOD2/C36-05 and RELAP5/MOD3 were implemented on the VPP500 computer system. (author)

NetMOD Version 2.0 User?s Manual.

Energy Technology Data Exchange (ETDEWEB)

Merchant, Bion J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-10-01

NetMOD ( Net work M onitoring for O ptimal D etection) is a Java-based software package for conducting simulation of seismic, hydracoustic, and infrasonic networks. Specifically, NetMOD simulates the detection capabilities of monitoring networks. Network simulations have long been used to study network resilience to station outages and to determine where additional stations are needed to reduce monitoring thresholds. NetMOD makes use of geophysical models to determine the source characteristics, signal attenuation along the path between the source and station, and the performance and noise properties of the station. These geophysical models are combined to simulate the relative amplitudes of signal and noise that are observed at each of the stations. From these signal-to-noise ratios (SNR), the probability of detection can be computed given a detection threshold. This manual describes how to configure and operate NetMOD to perform detection simulations. In addition, NetMOD is distributed with simulation datasets for the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO) International Monitoring System (IMS) seismic, hydroacoustic, and infrasonic networks for the purpose of demonstrating NetMOD's capabilities and providing user training. The tutorial sections of this manual use this dataset when describing how to perform the steps involved when running a simulation. ACKNOWLEDGEMENTS We would like to thank the reviewers of this document for their contributions.

NetMOD Version 2.0 Mathematical Framework

Energy Technology Data Exchange (ETDEWEB)

Merchant, Bion J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Young, Christopher J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Chael, Eric P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-08-01

NetMOD ( Net work M onitoring for O ptimal D etection) is a Java-based software package for conducting simulation of seismic, hydroacoustic and infrasonic networks. Network simulations have long been used to study network resilience to station outages and to determine where additional stations are needed to reduce monitoring thresholds. NetMOD makes use of geophysical models to determine the source characteristics, signal attenuation along the path between the source and station, and the performance and noise properties of the station. These geophysical models are combined to simulate the relative amplitudes of signal and noise that are observed at each of the stations. From these signal-to-noise ratios (SNR), the probabilities of signal detection at each station and event detection across the network of stations can be computed given a detection threshold. The purpose of this document is to clearly and comprehensively present the mathematical framework used by NetMOD, the software package developed by Sandia National Laboratories to assess the monitoring capability of ground-based sensor networks. Many of the NetMOD equations used for simulations are inherited from the NetSim network capability assessment package developed in the late 1980s by SAIC (Sereno et al., 1990).

RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Supplement 1, RELAP4/MOD5, Update 2

International Nuclear Information System (INIS)

Bruch, C.G.

1976-08-01

RELAP4/MOD5, Update 1 was released to the Nuclear Regulatory Commission in January 1976. Six months of extensive use of Update 1 has led to the identification and correction of a number of errors in the code, as well as some logic changes, additional Evaluation Model (EM) checks, and one model revision. These changes have culminated in the release of an improved code identified as RELAP4/MOD5, Update 2. The RELAP4/MOD5 interim User's Manual (Interim Report SRD-113-76) reflected the Update 1 version of the code. The purpose of the supplement presented is to update the Interim User's Manual for use with RELAP4/MOD5, Update 2. Major differences between Updates 1 and 2 and the checkout of Update 2 are discussed. The final version of the User's Manual will be written in accordance with Update 2 and will be published as ANCR-NUREG 1335 during September 1976. Annotation of the existing three volumes of the User's Manual to cross-reference this Supplement is recommended

TRAC-PF1/MOD2 status and plans

International Nuclear Information System (INIS)

Spore, J.W.; Steinke, R.G.; Nelson, R.A.; Cappiello, M.W.; Jenks, R.

1989-01-01

The development of the TRAC-PF1/MOD1 code was completed in July 1988 with the release of Version 14.4. A TRAC-PF1/MOD2 code development plan addresses code deficiencies identified in the MOD1 code in order to provide an accurate and defensible tool that can be used to simulate large-break loss-of-coolant accidents (LOCAs), small-break LOCAs, and operational transients. The MOD2 code development plan is an international cooperative effort that includes contributions from Los Alamos National Laboratory, Idaho National Engineering Laboratory (INEL), Japanese Atomic Energy Research Institute (JAERI), Cray Research, Central Electricity Generating Board (CEGB), and United Kingdom Atomic Energy Authority (UKAEA)

TRAC-PF1/MOD1 computer code

International Nuclear Information System (INIS)

Liles, D.R.; Mahaffy, J.H.

1984-01-01

The TRAC-P1 program was designed primarily for the analysis of large-break loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs). Because of its versatility, however, it can be applied directly to many analyses ranging from blowdowns in simple pipes to integral LOCA tests in multiloop facilities. A refined version, called TRAC-P1A, was released to the National Energy Software Center (NESC) in March 1979. Although it still treats the same class of problems, TRAC-P1A is more efficient than TRAC-P1 and incorporates improved hydrodynamic and heat-transfer models. It also is easier to implement on various computers. TRAC-PD2 contains improved reflood and heat-transfer models and improvements in the numerical solution methods. Although a large LOCA code, it has been applied successfully to small-break problems and to the Three Mile Island incident. Distinguishing characteristics of the TRAC-PF1/MOD1 are summarized

PL-MOD: a computer code for modular fault tree analysis and evaluation

International Nuclear Information System (INIS)

Olmos, J.; Wolf, L.

1978-01-01

The computer code PL-MOD has been developed to implement the modular methodology to fault tree analysis. In the modular approach, fault tree structures are characterized by recursively relating the top tree event to all basic event inputs through a set of equations, each defining an independent modular event for the tree. The advantages of tree modularization lie in that it is a more compact representation than the minimal cut-set description and in that it is well suited for fault tree quantification because of its recursive form. In its present version, PL-MOD modularizes fault trees and evaluates top and intermediate event failure probabilities, as well as basic component and modular event importance measures, in a very efficient way. Thus, its execution time for the modularization and quantification of a PWR High Pressure Injection System reduced fault tree was 25 times faster than that necessary to generate its equivalent minimal cut-set description using the computer code MOCUS

Assessment of RELAP5/MOD2 against a natural circulation experiment in Nuclear Power Plant Borssele

International Nuclear Information System (INIS)

Winters, L.

1993-07-01

As part of the ICAP (International Code Assessment and Applications Program) agreement between ECN (Netherlands Energy Research Foundation) and USNRC, ECN has performed a number of assessment calculations for the thermohydraulic system analysis code RELAP5/MOD2/36.05. This document describes the assessment of this computer program versus a natural circulation experiment as conducted at the Borssele Nuclear Power Plant. The results of this comparison show that the code RELAP5/MOD2 predicts well the natural circulation behaviour of Nuclear Power Plant Borssele

RELAP5/MOD2 Overview and Developmental. Assessment Results from TMl-1 Plant Transient Analysis

International Nuclear Information System (INIS)

Lin, J. C.; Tsai, C. C.; Ransom, V. H.; Johnsen, G. W.

2013-01-01

RELAP5/MOD2 is a new version of the RELAP5 thermal-hydraulic computer code containing improved modeling features that provide a generic capability for pressurized water reactor transient simulation. The objective of this paper is to provide code users with an overview of the code and to report developmental assessment results obtained from a Three Mile Island Unit One plant transient analysis. The assessment shows that the injection of highly sub-cooled water into a high-pressure primary coolant system does not cause unphysical results or pose a problem for RELAP5/MOD2. (author)

TRAC-PF1/MOD2 best-estimate analysis of a large-break LOCA in a 15 x 15 generic four-loop Westinghouse nuclear power plant

International Nuclear Information System (INIS)

Spore, J.W.; Lin, J.C.; Schnurr, N.M.; White, J.R.; Cappiello, M.C.

1992-01-01

Calculations of a large-break loss-of-coolant accident (LOCA) in a 15 x 15 generic four-loop Westinghouse nuclear power plant with both the TRAC-PF1/MOD1 and TRAC-PF1/MOD2 computer codes will be presented. The Transient Reactor Analysis Code (TRAC) has been developed by Los Alamos National Laboratory to provide advanced best-estimate simulations of real postulated transients in pressurized light-water reactors (LWRs) and for many related thermal-hydraulic facilities. The latest released version of TRAC is TRAC-PF1/MOD2. Significant improvements and enhancements over the MOD1 version were implemented in the MOD2 heat-transfer and constitutive models. One of the most significant improvements in the MOD2 code has been the implementation of the two-step numerics method in the three-dimensional components, which can significantly reduce run times for long, slow transients. A very important area of improvement has been in the reflood heat-transfer models. Developmental assessment results (i.e., code comparisons with experimental data) will be discussed for several separate-effects and integral test, including analysis of the Upper Plenum Test Facility (UPTF), the Cylindrical Core Test Facility (CCTF), and the Loss-of-Fluid Test Facility (LOFT). The assessment results provide information on the anticipated accuracy for the best-estimate models in the MOD2 computer code. The MOD1 to MOD2 comparison will provide an estimate for the effect of improved heat-transfer models on predicted peak cladding temperatures

Study of the Relap5/mod3.2 wall heat flux partitioning model

International Nuclear Information System (INIS)

Hari, S.; Hassan, Y.A.

2001-01-01

The performance of the subcooled boiling model adapted in RELAP5/MOD3.2 computer code has been assessed in detail for low-pressure conditions and it has been found that the void fraction profile is under-predicted. In general, any subcooled boiling model is composed of individual sub-models that account for the different physical mechanism that govern the overall process, as the wall vapor generation, interfacial shear and condensation etc. The wall heat flux partitioning model is one of the important sub-models that is a constituent of any subcooled boiling model. The function of this model is to apportion the wall heat flux to the different components (as the single/two phase fluid or bubble), as the case may be, in a two-phase flow-boiling scenario adjacent to a heated wall. The ''pumping factor'' approach is generally followed by most of the wall heat flux partitioning models, for partitioning the wall heat flux. In this work, the wall heat flux partitioning model of RELAP5/MOD3.2 computer code is studied; in particular, the ''pumping factor'' formulation in the present code version is assessed for its performance under low-pressure conditions. In addition, three different ''pumping factor'' formulations available in the literature have been introduced into the RELAP5/MOD3.2 code. Simulations of two low-pressure subcooled flow boiling experiments were performed with the refined code versions to determine the appropriate pumping factor to be used under these conditions. (author)

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

MOD-2 wind turbine system concept and preliminary design report. Volume 2: Detailed report

Science.gov (United States)

1979-01-01

The configuration development of the MOD-2 wind turbine system (WTS) is documented. The MOD-2 WTS project is a continuation of DOE programs to develop and achieve early commercialization of wind energy. The MOD-2 is design optimized for commercial production rates which, in multiunit installations, will be integrated into a utility power grid and achieve a cost of electricity at less than four cents per kilowatt hour.

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)

Relap4/SAS/Mod5 - A version of Relap4/Mod 5 adapted to IPEN/CNEN - SP computer center

International Nuclear Information System (INIS)

Sabundjian, G.

1988-04-01

In order to improve the safety of nuclear reactor power plants several computer codes have been developed in the area of thermal - hydraulics accident analysis. Among the public-available codes, RELAP4, developed by Aerojet Nuclear Company, has been the most popular one. RELAP4 has produced satisfactory results when compared to most of the available experimental data. The purposes of the present work are: optimization of RELAP4 output and messages by writing there information in temporary records, - display of RELAP4 results in graphical form through the printer. The sample problem consists on a simplified model of a 150 MW (e) PWR whose primary circuit is simulated by 6 volumes, 8 junctions and 1 heat slab. This new version of RELAP4 (named RELAP4/SAS/MOD5) have produced results which show that the above mentioned purposes have been reached. Obviously the graphical output by RELAP4/SAS/MOD5 favors the interpretation of results by the user. (author) [pt

RELAP5/MOD2: for PWR transient analysis

International Nuclear Information System (INIS)

Ransom, V.H.

1983-01-01

RELAP5 is a light water reactor system transient simulation code for use in nuclear plant safety analysis. Development of a new version, RELAP5/MOD2, has been completed and will be released to the United States Nuclear Regulatory Commission during September of 1983. The new and improved modeling capability of RELAP5/MOD2 is described and some developmental assessment results are presented. The future plans for extension to severe accident modeling are briefly discussed

RELAP5/MOD2 code modifications to obtain better predictions for the once-through steam generator

International Nuclear Information System (INIS)

Blanchat, T.; Hassan, Y.

1989-01-01

The steam generator is a major component in pressurized water reactors. Predicting the response of a steam generator during both steady-state and transient conditions is essential in studying the thermal-hydraulic behavior of a nuclear reactor coolant system. Therefore, many analytical and experimental efforts have been performed to investigate the thermal-hydraulic behavior of the steam generators during operational and accident transients. The objective of this study is to predict the behavior of the secondary side of the once-through steam generator (OTSG) using the RELAP5/MOD2 computer code. Steady-state conditions were predicted with the current version of the RELAP5/MOD2 code and compared with experimental plant data. The code predictions consistently underpredict the degree of superheat. A new interface friction model has been implemented in a modified version of RELAP5/MOD2. This modification, along with changes to the flow regime transition criteria and the heat transfer correlations, correctly predicts the degree of superheat and matches plant data

Post-test analysis of LOBI BT-01 using RELAP5/MOD2 and RELAP5/MOD3

International Nuclear Information System (INIS)

Holmes, B.J.

1991-08-01

LOBI is a high pressure, electrically heated integral system test facility simulating a KWU 1300 MW PWR scaled 1:712 by volume, although full scale has been maintained in the vertical direction. This report describes the results of an analysis of test BT-01, which simulates a 10% steam line break. The bulk of the analysis was performed using the Project Version of RELAP5/MOD2, with additional calculations using RELAP5/MOD3 for comparison. The codes provided generally good agreement with data. In particular, the break flows were well modelled, although the mass flow data proved to be unreliable, and this conclusion had to be derived from interpreting other signals. RELAP over-predicted primary/secondary heat transfer in the broken loop, however, leading to a more rapid cool-down of the primary circuit. Furthermore, the primary side pressure response was critically dependent upon the pressuriser behaviour, and the correct timing of the uncovery of the surge line. Inter-phase drag was not well predicted in the broken loop steam generator intermals, although some improvement was seen in the RELAP5/MOD3 predictions. MOD3 gave a reduction in primary/secondary heat transfer during the test pre-conditioning phase, resulting in a lower secondary side pressure at the start of the transient compared with MOD2. (author)

Simulation of single phase instability behaviour in a rectangular natural circulation loop using RELAP5/ MOD 3.2 computer code

International Nuclear Information System (INIS)

Sharma, Manish; Pilkhwal, D.S.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

2002-06-01

Occurrence of instability in natural circulation loops can lead to problems in control and occurrence of critical heat flux (CHF) during low flow periods. Remaining within an identified stable zone operation is therefore desirable. Natural circulation loops can pass through an unstable zone during start-up and power raising. In the present work RELAPS / MOD 3.2 computer code has been used to simulate the unstable oscillatory behavior observed in a rectangular natural circulation loop having horizontal heater and horizontal cooler (HHHC) orientation. The results were compared with the experimental data. This report describes the nodalization scheme adopted tor this work and results of the analysis in detail. (author)

Assessment and improvement of condensation models in RELAP5/MOD3.2

Energy Technology Data Exchange (ETDEWEB)

Choi, Ki Yong; Park, Hyun Sik; Kim, Sang Jae; No, Hee Chen [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

The condensation models in the standard RELAP5/MOD3.2 code are assessed and improved based on the database, which is constructed from the previous experimental data on various condensation phenomena. The default model of the laminar film condensation in RELAP5/MOD3.2 does not give any reliable predictions, and its alternative model always predicts higher values than the experimental data. Therefore, it is needed to develop a new correlation based on the experimental data of various operating ranges in the constructed database. The Shah correlation, which is used to calculate the turbulent film condensation heat transfer coefficients in the standard RELAP5/MOD3.2, well predicts the experimental data in the database. The horizontally stratified condensation model of RELAP5/MOD3.2 overpredicts both cocurrent and countercurrent experimental data. The correlation proposed by H.J.Kim predicts the database relatively well compared with that of RELAP6/MOD3.2. The RELAP5/MOD3.2 model should use the liquid velocity for the calculation of the liquid Reynolds number and be modified to consider the effects of the gas velocity and the film thickness. 2 refs., 5 figs., 1 tab. (Author)

Assessment and improvement of condensation models in RELAP5/MOD3.2

Energy Technology Data Exchange (ETDEWEB)

Choi, Ki Yong; Park, Hyun Sik; Kim, Sang Jae; No, Hee Chen [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1997-12-31

The condensation models in the standard RELAP5/MOD3.2 code are assessed and improved based on the database, which is constructed from the previous experimental data on various condensation phenomena. The default model of the laminar film condensation in RELAP5/MOD3.2 does not give any reliable predictions, and its alternative model always predicts higher values than the experimental data. Therefore, it is needed to develop a new correlation based on the experimental data of various operating ranges in the constructed database. The Shah correlation, which is used to calculate the turbulent film condensation heat transfer coefficients in the standard RELAP5/MOD3.2, well predicts the experimental data in the database. The horizontally stratified condensation model of RELAP5/MOD3.2 overpredicts both cocurrent and countercurrent experimental data. The correlation proposed by H.J.Kim predicts the database relatively well compared with that of RELAP6/MOD3.2. The RELAP5/MOD3.2 model should use the liquid velocity for the calculation of the liquid Reynolds number and be modified to consider the effects of the gas velocity and the film thickness. 2 refs., 5 figs., 1 tab. (Author)

Installation and validation of RELAP4/MOD6 on the VAX 11/780 computer at IKE

International Nuclear Information System (INIS)

Lang, U.; Haussmann, C.

1983-06-01

RELAP4/MOD6 is a FORTRAN 4 code for the transient thermohydraulic analysis of nuclear reactors and similar systems. This code has been developed by the Idaho National Engeneering Laboratory for CDC and IBM computers. The implementation of the code on a VAX 11/780 has been possible due to the fact, that this computer is a byte oriented system as the IBM machines, and that the code has been wirtten only in part in a machine dependent way. Two versions of RELAP4/MOD6 are available on the VAX-system, with different dimensions for the input parameters (SI-units or BTU-units). The implementation of the two versions and their validation is described in this report. (orig.) [de

CORCON-MOD1 modelling improvements

International Nuclear Information System (INIS)

Corradini, M.L.; Gonzales, F.G.; Vandervort, C.L.

1986-01-01

Given the unlikely occurrence of a severe accident in a light water reactor (LWR), the core may melt and slump into the reactor cavity below the reactor vessel. The interaction of the molten core with exposed concrete (a molten-core-concrete-interaction, MCCI) causes copious gas production which influences further heat transfer and concrete attack and may threaten containment integrity. In this paper the authors focus on the low-temperature phase of the MCCI where the molten pool is partially solidified, but is still capable of attacking concrete. The authors have developed some improved phenomenological models for pool freezing and molten core-coolant heat transfer and have incorporated them into the CORCON-MOD1 computer program. In the paper the authors compare the UW-CORCON/MOD1 calculations to CORCON/MOD2 and WECHSL results as well as the BETA experiments which are being conducted in Germany

RELAP5/MOD2 calculation of OECD LOFT test LP-FW-01

International Nuclear Information System (INIS)

Croxfod, M.G.; Harwood, C.; Hall, P.C.

1992-04-01

RELAP5/MOD2 is being used by GDCD for calculation of certain small break loss-of-coolant accidents and pressurized transients in the Sizewell ''B'' PWR. To test the ability of RELAP5/MOD2 to model the primary feed-and-bleed recovery procedure following a complete loss- of-feedwater event, post test calculations have been carried out of OECD LOFT test LP-FW-01. This report describes the comparison between the code calculations and the test data. It is found that although the standard version of RELAP5/MOD2 gives a reasonable prediction of the experimental transient, the long term pressure history is better calculated with a modified code version containing a revised horizontal stratification entrainment model. The latter allows an improved calculation of entrainment of liquid from the hot leg into the surge line. RELAP5/MOD2 is found to give a more accurate simulation of the experimental transient than was achieved in previous UK studies using RETRAN-02/MOD2

THYDE-B1/MOD2: a computer code for analysis of small-break loss-of-coolant accidents of boiling water reactors

International Nuclear Information System (INIS)

Nakamura, Hideo; Muramatsu, Ken; Kukita, Yutaka; Tasaka, Kanji

1988-04-01

THYDE-B1/MOD2 is a fast-running best estimate (BE) computer code to analyze thermal-hydraulic behaviors of the reactor cooling system of a boiling water reactor (BWR), mainly, during a small-break loss-of-coolant accident (SBLOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions consist of subcooled liquid, saturated mixture and saturated steam regions from the volume bottom. The regions are separated by two horizontal moving boundaries which are tracked by mass and energy balances for each region. With this three region node model, the interior of the pressure vessel can be represented by only two volumes: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous node model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SBLOCAs in which the thermal-hydraulic behavior is relatively slow and gravity controlled. The code has been improved and modified from the last version of the code, THYDE-B1/MOD1, especially in the phase separation model which is used in the mixture level calculation in the three region node model. Then, a good predictability of the code has been indicated through the comparison of calculated results with various SBLOCA test data including ROSA-III of JAERI and FIST of the General Electric Co. This report presents the code modifications and input data requirements of the THYDE-B1/MOD2 code. (author)

Automotive Stirling engine: Mod 2 design report

Science.gov (United States)

Nightingale, Noel P.

1986-01-01

The design of an automotive Stirling engine that achieves the superior fuel economy potential of the Stirling cycle is described. As the culmination of a 9-yr development program, this engine, designated the Mod 2, also nullifies arguments that Stirling engines are heavy, expensive, unreliable, demonstrating poor performance. Installed in a General Motors Chevrolet Celebrity car, this engine has a predicted combined fuel economy on unleaded gasoline of 17.5 km/l (41 mpg)- a value 50% above the current vehicle fleet average. The Mod 2 Stirling engine is a four-cylinder V-drive design with a single crankshaft. The engine is also equipped with all the controls and auxiliaries necessary for automotive operation.

Comparision of calculations for the ROSA-IV LSTF with RELAP5/MOD0 and RELAP5/MOD1 (cycle 1)

International Nuclear Information System (INIS)

Fineman, C.P.; Tanaka, Mitsugu; Tasaka, Kanji

1982-03-01

10% and 2.5% cold leg break analyses have been completed for the ROSA-IV Large Scale Test Facility (LSTF) with the RELAP5/MOD0 and RELAP5/MOD1, cycle 1, computer codes. Comparisons between the calculations were made to determine any differences in the results obtained from the two versions of RELAP5. Differences in the two calculations were found which can be attributed to changes in the flow regime maps and critical flow model. (author)

Una riconsiderazione sul ripostiglio di S'Adde 'e S'Ulumu-Usini

Directory of Open Access Journals (Sweden)

Salvatore Merella

2012-12-01

Full Text Available In this poster the author presents the results of a study on several pieces of ashlar masonry were recently discovered in a locale called S'Iscia 'e Su Puttu, indicating the likely presence of sacred place belonging to the Nuragic - Bronze and Early Iron Age - phase. It is located very close to the site of S'Adde 'e S'Ulumu, where an important hoard of bronze objects was found, once considered to belong to a single individual for the absence of any clear archaeological context. This hoard acquires a new dimension thanks to the new data presented here: it could have reasonably been part of a set of ritual activities carried out in a sacred area formed of buildings and spaces between them, a common pattern in Sardinian communities during the Early Iron phase. This poster was presented at the Workshop Materiali e contesti dell'Età del Ferro sarda (Materials and contexts in the Sardinian Iron Age, organised by the University of Glasgow and the Comune di San Vero Milis on the 25th of May 2012, and supported for by the Royal Society of Edinburgh, the Comune di San Vero Milis, the University of Glasgow and the Carnegie Trust of the Universities of Scotland.

RELAP5/MOD2 models and correlations

International Nuclear Information System (INIS)

Dimenna, R.A.; Larson, J.R.; Johnson, R.W.; Larson, T.K.; Miller, C.S.; Streit, J.E.; Hanson, R.G.; Kiser, D.M.

1988-08-01

A review of the RELAP5/MOD2 computer code has been performed to assess the basis for the models and correlations comprising the code. The review has included verification of the original data base, including thermodynamic, thermal-hydraulic, and geothermal conditions; simplifying assumptions in implementation or application; and accuracy of implementation compared to documented descriptions of each of the models. An effort has been made to provide the reader with an understanding of what is in the code and why it is there and to provide enough information that an analyst can assess the impact of the correlation or model on the ability of the code to represent the physics of a reactor transient. Where assessment of the implemented versions of the models or correlations has been accomplished and published, the assessment results have been included

Validation of One-Dimensional Module of MARS-KS1.2 Computer Code By Comparison with the RELAP5/MOD3.3/patch3 Developmental Assessment Results

International Nuclear Information System (INIS)

Bae, S. W.; Chung, B. D.

2010-07-01

This report records the results of the code validation for the one-dimensional module of the MARS-KS thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 Code Developmental Assessment Problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The result suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

Double blind post-test prediction for LOBI-MOD2 small break experiment A2-81 using RELAP5/MOD1/19 computer code as contribution to international CSNI-standardproblem no. 18

International Nuclear Information System (INIS)

Jacobs, G.; Mansoor, S.H.

1986-06-01

The first small break experiment A2-81 performed in the LOBI-MOD2 test facility was the base of the 18th international CSNI standard problem (ISP 18). Taking part in this exercise, a blind post-test prediction was performed using the light water reactor transient analysis code RELAP5/MOD1. This paper describes the input model preparation and summarizes the findings of the pre-calculation comparing the calculational results with the experimental data. The results show that there was a good agreement between prediction and experiment in the initial stage (up to 250 sec) of the transient and an adequate prediction of the global behaviour (thermal response of the core), which is important for safety related considerations. However, the prediction confirmed some deficiencies of the models in the code concerning vertical and horizontal stratification resulting in a high break mass flow and an erroneous distribution of mass over the primary loops. (orig.) [de

A study of the dispersed flow interfacial heat transfer model of RELAP5/MOD2.5 and RELAP5/MOD3

Energy Technology Data Exchange (ETDEWEB)

Andreani, M. [Swiss Federal Institute of Technology, Zurich (Switzerland); Analytis, G.T.; Aksan, S.N. [Paul Scherrer Institute, Villigen (Switzerland)

1995-09-01

The model of interfacial heat transfer for the dispersed flow regime used in the RELAP5 computer codes is investigated in the present paper. Short-transient calculations of two low flooding rate tube reflooding experiments have been performed, where the hydraulic conditions and the heat input to the vapour in the post-dryout region were controlled for the predetermined position of the quench front. Both RELAP5/MOD2.5 and RELAP5/MOD3 substantially underpredicted the exit vapour temperature. The mass flow rate and quality, however, were correct and the heat input to the vapour was larger than the actual one. As the vapour superheat at the tube exit depends on the balance between the heat input from the wall and the heat exchange with the droplets, the discrepancy between the calculated and the measured exit vapour temperature suggested that the inability of both codes to predict the vapour superheat in the dispersed flow region is due to the overprediction of the interfacial heat transfer rate.

BEACON/MOD2A: a computer program for subcompartment analysis of nuclear reactor containment. A user's manual

International Nuclear Information System (INIS)

Wells, R.A.

1979-03-01

The BEACON code is a Best Estimate Advanced Containment code which being developed by EG and G, Idaho, Inc., at the Idaho National Engineering Laboratory. The program is designed to perform a best estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped-parameter representations for the various parts of the system. The current version of BEACON, which is designated BEACON/MOD2A, contains mass and heat transfer models for wall film and for wall conduction. It is suitable for the evaluation of short term transients in PWR dry containment systems. This manual describes the models employed in BEACON/MOD2A and specifies code implementation requirements. It provides application information for input data preparation and for output data interpretation

Comparison of MAAP4.03 with RELAP/SCDAPSIM/MOD3.2

Energy Technology Data Exchange (ETDEWEB)

Kohriyama, T.; Ohtani, M. [Technical Support Project, Institute of Nuclear Technology, Institute of Nuclear Safety System, Incorporated, Mihama, Fukui (Japan); Ezzidi, A.; Morota, H. [Computer Software Development Co., Ltd., Tokyo (Japan)

2000-11-01

The MAAP4.03 code has been widely used for analyses of severe accident phenomena. It is, however, a system level code applying simpler models and could show phenomenological uncertainties. In order to support MAAP4.03 by a more detailed mechanistic code such as RELAP/SCDAPSIM/MOD3.2, code-to-code comparisons are performed. For a typical 4 loop PWR, analyses of two severe accident sequences were executed. Both codes predicted similar tendencies until the beginning of core melt. MAAP4.03 showed earlier slumping of molten core to a lower head of a reactor pressure vessel than RELAP/SCDAPSIM/MOD3.2. MAAP4.03 considers only axial flows of molten core and crusts suddenly breach by Creep Rupture. RELAP/SCDAPSIM/MOD3.2 treats axial and radial spreads by repeated cycles of melting, flowing and freezing. Bottom crusts can be supported by intact fuel rods. By these more realistic RELAP/SCDAPSIM/MOD3.2 models, MAAP4.03 could be supported maintaining conservatism. (author)

Code development and analysis program. RELAP4/MOD7 (Version 2): user's manual

International Nuclear Information System (INIS)

1978-08-01

This manual describes RELAP4/MOD7 (Version 2), which is the latest version of the RELAP4 LPWR blowdown code. Version 2 is a precursor to the final version of RELAP4/MOD7, which will address LPWR LOCA analysis in integral fashion (i.e., blowdown, refill, and reflood in continuous fashion). This manual describes the new code models and provides application information required to utilize the code. It must be used in conjunction with the RELAP4/MOD5 User's Manual (ANCR-NUREG-1335, dated September 1976), and the RELAP4/MOD6 User's Manual

Assessment of RELAP/MOD2 using large break loss-of-coolant experimental data

International Nuclear Information System (INIS)

Kao, L.; Liao, L.Y.; Liang, K.S.; Wang, S.F.; Chen, Y.B.

1989-01-01

In this paper assessment of RELAP5/MOD2 using LOFT L2-5 and Semiscale S-06-3 tests are performed to provide information of the code capability and its limitation in analyzing large break LOCA of a nuclear power plant. Experiments L2-5 and S-06-3 are conducted to simulate a hypothetical LOCA which results from a 200% double-ended offset shear break in the cold-leg of a typical pressurized water reactor by utilizing scaling facilities of the LOFT and Semiscale Mod-1 systems, respectively. The RELAP5/MOD2 calculations for both tests begin with break initiation and subsequent blowdown, continue through lower plenum refill, core reflood, and terminate with corewide quench. Major phenomena of both large break loss-of-coolant tests are well predicted by RELAP5/MOD2. The results indicate that the break flow and system pressure are reasonably calculated. The cladding temperature response during blowdown period, which is the major importance to a large break LOCA, calculated by RELAP5/MOD2 shows good agreement with the test data

Integrated uncertainty analysis using RELAP/SCDAPSIM/MOD4.0

International Nuclear Information System (INIS)

Perez, M.; Reventos, F.; Wagner, R.; Allison, C.

2009-01-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis package being developed jointly by the Technical University of Catalunya (UPC) and Innovative Systems Software (ISS). The integrated uncertainty analysis approach used in the package uses the following steps: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. The first four steps are performed by the user prior to the RELAP/SCDAPSIM/MOD4.0 analysis. The remaining steps are included with the MOD4.0 integrated uncertainty analysis (IUA) package. This paper briefly describes the integrated uncertainty analysis package including (a) the features of the package, (b) the implementation of the package into RELAP/SCDAPSIM/MOD4.0, and

Mod two homology and cohomology

CERN Document Server

Hausmann, Jean-Claude

2014-01-01

Cohomology and homology modulo 2 helps the reader grasp more readily the basics of a major tool in algebraic topology. Compared to a more general approach to (co)homology this refreshing approach has many pedagogical advantages: It leads more quickly to the essentials of the subject, An absence of signs and orientation considerations simplifies the theory, Computations and advanced applications can be presented at an earlier stage, Simple geometrical interpretations of (co)chains. Mod 2 (co)homology was developed in the first quarter of the twentieth century as an alternative to integral homology, before both became particular cases of (co)homology with arbitrary coefficients. The first chapters of this book may serve as a basis for a graduate-level introductory course to (co)homology. Simplicial and singular mod 2 (co)homology are introduced, with their products and Steenrod squares, as well as equivariant cohomology. Classical applications include Brouwer's fixed point theorem, Poincaré duality, Borsuk-Ula...

Vectorization and improvement of nuclear codes (MEUDAS4, FORCE, STREAM V2.6, HEATING7-VP, SCDAP/RELAP5/MOD2.5, NBI3DGFN)

International Nuclear Information System (INIS)

Nemoto, Toshiyuki; Suzuki, Koichiro; Isobe, Nobuo; Machida, Masahiko; Osanai, Seiji; Yokokawa, Mitsuo

1992-09-01

Eight nuclear codes have been vectorized and modified to improve their performance. These codes are magnetic fluid equilibrium code MEUDAS4 (CR and FFT versions), the magnetic field analysis code FORCE, the three-dimensional heat fluid analysis code STREAM V2.6, the three-dimensional heat analysis code HEATING 7-VP, the severe accident transient analysis code SCDAP/RELAP 5/MOD 2.5 for light water reactors, the ion beam orbital analysis code NBI3DGFN, and a free electron laser analysis code. The speedup ratios of the vectorized versions to the original ones in scalar mode are 2.3-4.9, 1.9-5.4, 2.6-6.2, and 1.9 for the MEUDAS4, STREAM, FORCE, and free electron laser analysis code, respectively. The definition method of the computational regions in the HEATING7-VP is improved. The SCDAP/RELAP5/MOD2.5 is modified to use extended memory regions of the computer. In this report, outlines of the codes, techniques used in the vectorization and reorganization of the codes, verification of computed results, and improvement on the performance are presented. (author)

MOD-1 Wind Turbine Generator Analysis and Design Report, Volume 2

Science.gov (United States)

1979-01-01

The MOD-1 detail design is appended. The supporting analyses presented include a parametric system trade study, a verification of the computer codes used for rotor loads analysis, a metal blade study, and a definition of the design loads at each principal wind turbine generator interface for critical loading conditions. Shipping and assembly requirements, composite blade development, and electrical stability are also discussed.

Distribution of the type III DNA methyltransferases modA, modB and modD among Neisseria meningitidis genotypes: implications for gene regulation and virulence.

Science.gov (United States)

Tan, Aimee; Hill, Dorothea M C; Harrison, Odile B; Srikhanta, Yogitha N; Jennings, Michael P; Maiden, Martin C J; Seib, Kate L

2016-02-12

Neisseria meningitidis is a human-specific bacterium that varies in invasive potential. All meningococci are carried in the nasopharynx, and most genotypes are very infrequently associated with invasive meningococcal disease; however, those belonging to the 'hyperinvasive lineages' are more frequently associated with sepsis or meningitis. Genome content is highly conserved between carriage and disease isolates, and differential gene expression has been proposed as a major determinant of the hyperinvasive phenotype. Three phase variable DNA methyltransferases (ModA, ModB and ModD), which mediate epigenetic regulation of distinct phase variable regulons (phasevarions), have been identified in N. meningitidis. Each mod gene has distinct alleles, defined by their Mod DNA recognition domain, and these target and methylate different DNA sequences, thereby regulating distinct gene sets. Here 211 meningococcal carriage and >1,400 disease isolates were surveyed for the distribution of meningococcal mod alleles. While modA11-12 and modB1-2 were found in most isolates, rarer alleles (e.g., modA15, modB4, modD1-6) were specific to particular genotypes as defined by clonal complex. This suggests that phase variable Mod proteins may be associated with distinct phenotypes and hence invasive potential of N. meningitidis strains.

Qualification of the AUTOBUS Mod. 2 Code

International Nuclear Information System (INIS)

Ciarniello, U.; Peroni, P.

1988-01-01

The paper presents the qualification of AUTOBUS MOD.2 code. After a brief description of the code itself, all the critical experiments simulated by the code are illustrated to prove the accuracy of criticality calculation and power distribution. An interpretation of the results and a conclusion close this presentation

Comparison of Scrape-off Layer Turbulence in Alcator C-Mod with Three Dimensional Gyrofluid Computations

International Nuclear Information System (INIS)

Zweben, S.J.; Scott, B.D.; Terry, J.L.; LaBombard, B.; Hughes, J.W.; Stotler, D.P.

2009-01-01

This paper describes quantitative comparisons between turbulence measured in the scrape-off layer (SOL) of Alcator C-Mod (S. Scott, A. Bader, M. Bakhtiari et al., Nucl. Fusion 47, S598 (2007)) and three dimensional computations using electromagnetic gyrofluid equations in a two-dimensional tokamak geometry. These comparisons were made for the outer midplane SOL for a set of inner-wall limited, near-circular Ohmic plasmas. The B field and plasma density were varied to assess gyroradius and collisionality scaling. The poloidal and radial correlation lengths in the experiment and computation agreed to within a factor of 2 and did not vary significantly with either B or density. The radial and poloidal propagation speeds and the frequency spectra and poloidal k-spectra also agreed fairly well. However, the autocorrelation times and relative Da fluctuation levels were higher in the experiment by more than a factor of 2. Possible causes for these disagreements are discussed.

Independent assessment of TRAC-PD2 and RELAP5/MOD1 codes at BNL in FY 1981

International Nuclear Information System (INIS)

Saha, P.; Jo, J.H.; Neymotin, L.; Rohatgi, U.S.; Slovik, G.

1982-12-01

This report documents the independent assessment calculations performed with the TRAC-PD2 and RELAP/MOD1 codes at Brookhaven National Laboratory (BNL) during Fiscal Year 1981. A large variety of separate-effects experiments dealing with (1) steady-state and transient critical flow, (2) level swell, (3) flooding and entrainment, (4) steady-state flow boiling, (5) integral economizer once-through steam generator (IEOTSG) performance, (6) bottom reflood, and (7) two-dimensional phase separation of two-phase mixtures were simulated with TRAC-PD2. In addition, the early part of an overcooling transient which occurred at the Rancho Seco nuclear power plant on March 20, 1978 was also computed with an updated version of TRAC-PD2. Three separate-effects tests dealing with (1) transient critical flow, (2) steady-state flow boiling, and (3) IEOTSG performance were also simulated with RELAP5/MOD1 code. Comparisons between the code predictions and the test data are presented

The implementation of the CDC version of RELAP5/MOD1/019 on an IBM compatible computer system (AMDAHL 470/V8)

International Nuclear Information System (INIS)

Kolar, W.; Brewka, W.

1984-01-01

RELAP5/MOD1 is an advanced one-dimensional best estimate system code, which is used for safety analysis studies of nuclear pressurized water reactor systems and related integral and separate effect test facilities. The program predicts the system response for large break, small break LOCA and special transients. To a large extent RELAP5/MOD1 is written in Fortran, only a small part of the program is coded in CDC assembler. RELAP5/MOD1 was developed on the CDC CYBER 176 at INEL*. The code development team made use of CDC system programs like the CDC UPDATE facility and incorporated in the program special purpose software packages. The report describes the problems which have been encountered when implementing the CDC version of RELAP5/MOD1 on an IBM compatible computer systems (AMDAHL 470/V8)

Simulation of accident and restrained transients in PWR nuclear power plant with RELAP 5/MOD 1 computer code

International Nuclear Information System (INIS)

Silva Filho, E.

1986-01-01

The computer code RELAP5/MOD1 has been utilized to investigate the thermal-hydraulic behaviour of a standard 1300 Mwe pressurized water reactor plant of the KWU design during a station blackout and during a loss-of-coolant accident involving 2% break in the cross-sectional area the cold leg in one of the four loops and located between the pump and the reactor pressure vessel. During the simulations the reactor scram system and the emergency coolant system were considered inactive. (Author) [pt

The Escherichia coli modE gene: effect of modE mutations on molybdate dependent modA expression.

Science.gov (United States)

McNicholas, P M; Chiang, R C; Gunsalus, R P

1996-11-15

The Escherichia coli modABCD operon, which encodes a high-affinity molybdate uptake system, is transcriptionally regulated in response to molybdate availability by ModE. Here we describe a highly effective enrichment protocol, applicable to any gene with a repressor role, and establish its application in the isolation of transposon mutations in modE. In addition we show that disruption of the ModE C-terminus abolishes derepression in the absence of molybdate, implying this region of ModE controls the repressor activity. Finally, a mutational analysis of a proposed molybdate binding motif indicates that this motif does not function in regulating the repressor activity of ModE.

NAUA Mod 4

International Nuclear Information System (INIS)

Bunz, H.; Koyro, M.; Schoeck, W.

1983-08-01

This report describes the computer program NAUA Mod4. Its purpose is to calculate the behaviour of a polydisperse aerosol system in a closed vessel containing a condensing atmosphere as a function of the time. The main object is to explain the physical background and to describe the structure of the code and the input and output in detail. (orig.) [de

Mod-2 wind turbine system concept and preliminary design report. Volume 1: Executive summary

Science.gov (United States)

1979-01-01

The configuration development of the MOD-2 wind turbine system is presented. The MOD-2 is design optimized for commercial production rates which, in multi-unit installations, will be integrated into a utility power grid and achieve a cost of electricity at less than 4 cents per kilowatt hour.

CONTEMPT4/MOD2: a multicompartment containment system analysis program

International Nuclear Information System (INIS)

Metcalfe, L.J.; Mings, W.J.; Hartman, J.E.; Crail, A.C.

1978-02-01

CONTEMPT4/MOD2 is a digital computer program, written in FORTRAN IV, which describes the behavior of multicompartment pressurized water reactor (PWR) containment systems and experimental containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and intercompartment mass and energy exchange based on user-supplied values for compartment descriptions, time step and edit controls, and selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, fan coolers, heat conducting structures, sump drain, and PWR ice condensers. Dynamic storage allocations (DSA) is used to limit the amount of computer core used for each problem. Optional automatic time step control allows the code to determine time step sizes within limits dictated by the user. Multicompartment capability (up to 999 individual compartments) and generalized, user-oriented input data descriptions permit improved flexibility over previous codes in the CONTEMPT series. Analytical model descriptions, input instructions, and sample problem results are presented

BLOW.MOD2: program for a vessel depressurization calculation with the contribution of structures

International Nuclear Information System (INIS)

Doval, A.

1990-01-01

The BLOW.MOD2 program developed to calculate pressure vessels' depressurization is presented, considering heat contribution of the structures. The results are opposite to those obtained from other more complex numerical models, being the comparison extremely satisfactory. BLOW.MOD2 is a software of the 'Systems Sub-Branch', INVAP S.E. (Author) [es

RELAP5/MOD3 code coupling model

International Nuclear Information System (INIS)

Martin, R.P.; Johnsen, G.W.

1994-01-01

A new capability has been incorporated into RELAP5/MOD3 that enables the coupling of RELAP5/MOD3 to other computer codes. The new capability has been designed to support analysis of the new advanced reactor concepts. Its user features rely solely on new RELAP5 open-quotes styledclose quotes input and the Parallel Virtual Machine (PVM) software, which facilitates process management and distributed communication of multiprocess problems. RELAP5/MOD3 manages the input processing, communication instruction, process synchronization, and its own send and receive data processing. The flexible capability requires that an explicit coupling be established, which updates boundary conditions at discrete time intervals. Two test cases are presented that demonstrate the functionality, applicability, and issues involving use of this capability

TRAC-PF1/MOD1 computer code

International Nuclear Information System (INIS)

Liles, D.R.; Mahaffy, J.H.

1983-01-01

TRAC-PF1 was designed to improve the ability of TRAC-PD2 to handle small-break LOCAs and other transients. TRAC-PF1 has all of the major improvements of TRAC-PD2 but, in addition, uses a full two-fluid model with two-step numerics in the one-dimensional components. The two-fluid model, in conjunction with a stratified-flow regime, handles countercurrent flow better than the drift-flux model previously used. The two-step numerics allow large time steps to be taken for slow transients. TRAC-PF1/MOD1 was designed to provide full balance-of-plant modeling capabilities. This required addition of a general capability for modeling plant control systems. The steam generator model was replaced to allow a wider variety of feedwater connections and better modeling of steam tube ruptures. A special turbine component also has been added, but new components were not required for adequate modeling of condensors, heaters, and pumps in the secondary system

Progress on MOD/RABiTSTM 2G HTS wire

International Nuclear Information System (INIS)

Rupich, M.W.; Zhang, W.; Li, X.; Kodenkandath, T.; Verebelyi, D.T.; Schoop, U.; Thieme, C.; Teplitsky, M.; Lynch, J.; Nguyen, N.; Siegal, E.; Scudiere, J.; Maroni, V.; Venkataraman, K.; Miller, D.; Holesinger, T.G.

2004-01-01

The development of the second generation (2G) high temperature superconducting wire has advanced beyond initial laboratory demonstrations and is now focused on developing and testing high critical current conductor designs required for commercial applications. The approach pursued at American Superconductor for 2G wire manufacturing is based on the combination of the RABiTS TM substrate-buffer technology with metal organic deposition (MOD) of the YBCO layer. This MOD/RABiTS TM approach has been demonstrated in 10 m lengths with critical currents of up to 184 A/cm-width (∼2.3 MA/cm 2 ) and in short length with critical currents of up to 270 A/cm-width (∼3.4 MA/cm 2 ). In addition to a high critical current, the superconducting wire must also meet stringent mechanical and electrical stability requirements that vary by application. Commercially viable architectures designed to meet these specifications have been fabricated and tested. Wires manufactured by this process have been successfully tested in prototype cable and coil applications

Analysis of Rod Withdrawal at Power (RWAP) Accident using ATHLET Mod 2.2 Cycle A and RELAP5/mod 3.3 Codes

International Nuclear Information System (INIS)

Bencik, V.; Cavlina, N.; Grgic, D.

2012-01-01

The system code ATHLET is being developed at Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS) in Germany. In 1996, the NPP Krsko (NEK) input deck for ATHLET Mod 1.1 Cycle C has been developed at Faculty of Electrical Engineering (FER), University of Zagreb. The input deck was tested by analyzing the realistic plant event 'Main Steam Isolation Valve Closure' and the results were assessed against the measured data. The input deck was established before plant modernization that took place in 2000 and included the power uprate and SG replacement. The released ATHLET version (Mod 2.2 Cycle A) is now being available at FER Zagreb. Accordingly, the NEK input deck for ATHLET Mod 2.2 Cycle A has been developed. A completely new input deck has been created taking into account the large number of changes due to power uprate and SG replacement as well as taking advantage of developmental work on NEK data base performed at FER. The new NEK input deck for ATHLET code has been tested by analyzing the Rod Withdrawal Power (RWAP) accident and the results were assessed against the analysis performed by RELAP5/mod 3.3 code. The RWAP accident can be either Departure from Nucleate Boiling (DNB) ratio or overpower limiting accident depending on initial power and reactivity insertion rate. Since the automatic rod control system is assumed unavailable, the only negative reactivity is due to Doppler and moderator feedback. Consequently, the nuclear power and the transferred heat in the steam generators (SGs) increase. Since the steam flow to the turbine and the extracted power from the SGs remain constant, the SG secondary pressure and the temperatures on the primary side increase. Unless terminated by manual or automatic action, the power mismatch between primary and secondary side and the resultant coolant temperature rise could eventually result in DNB ratio and/or fuel centreline melt. In order to avoid core damage, the reactor protection system is designed to automatically

An analysis of the binding of repressor protein ModE to modABCD (molybdate transport) operator/promoter DNA of Escherichia coli.

Science.gov (United States)

Grunden, A M; Self, W T; Villain, M; Blalock, J E; Shanmugam, K T

1999-08-20

Expression of the modABCD operon in Escherichia coli, which codes for a molybdate-specific transporter, is repressed by ModE in vivo in a molybdate-dependent fashion. In vitro DNase I-footprinting experiments identified three distinct regions of protection by ModE-molybdate on the modA operator/promoter DNA, GTTATATT (-15 to -8; region 1), GCCTACAT (-4 to +4; region 2), and GTTACAT (+8 to +14; region 3). Within the three regions of the protected DNA, a pentamer sequence, TAYAT (Y = C or T), can be identified. DNA-electrophoretic mobility experiments showed that the protected regions 1 and 2 are essential for binding of ModE-molybdate to DNA, whereas the protected region 3 increases the affinity of the DNA to the repressor. The stoichiometry of this interaction was found to be two ModE-molybdate per modA operator DNA. ModE-molybdate at 5 nM completely protected the modABCD operator/promoter DNA from DNase I-catalyzed hydrolysis, whereas ModE alone failed to protect the DNA even at 100 nM. The apparent K(d) for the interaction between the modA operator DNA and ModE-molybdate was 0.3 nM, and the K(d) increased to 8 nM in the absence of molybdate. Among the various oxyanions tested, only tungstate replaced molybdate in the repression of modA by ModE, but the affinity of ModE-tungstate for modABCD operator DNA was 6 times lower than with ModE-molybdate. A mutant ModE(T125I) protein, which repressed modA-lac even in the absence of molybdate, protected the same region of modA operator DNA in the absence of molybdate. The apparent K(d) for the interaction between modA operator DNA and ModE(T125I) was 3 nM in the presence of molybdate and 4 nM without molybdate. The binding of molybdate to ModE resulted in a decrease in fluorescence emission, indicating a conformational change of the protein upon molybdate binding. The fluorescence emission spectra of mutant ModE proteins, ModE(T125I) and ModE(Q216*), were unaffected by molybdate. The molybdate-independent mutant Mod

Mod i ledelse

DEFF Research Database (Denmark)

Mellon, Karsten

2016-01-01

Mod i ledelse er en efterspurgt vare i offentligt regi som modsvar på stigende kompleksitet og pres. Men hvad er ’mod i ledelse’ – og er du selv en modig leder?......Mod i ledelse er en efterspurgt vare i offentligt regi som modsvar på stigende kompleksitet og pres. Men hvad er ’mod i ledelse’ – og er du selv en modig leder?...

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.

The alcator C-MOD control system

International Nuclear Information System (INIS)

Bosco, J.; Fairfax, S.

1992-01-01

The Alcator C-MOD experiment includes over 30 engineering and diagnostic subsystems. The control system hardware and software is a mixture of custom and commercial products which includes sensors, signal conditioners, hard-wired controls, programmable logic controllers, displays, a hybrid analog/digital computer, networked personal computers, and networked VAX workstations. This paper describes the computer-based portions of the control system. The control system coordinates all C-MOD systems including power, vacuum, heating and cooling, access control, plasma shape and position control, and diagnostics. Programmable logic controllers (PLC's) are located near each subsystem. The control room is isolated by fiber optics. Functions that are essential to personnel or equipment safety (e.g. access control) are implemented in hardwired logic and monitored but not controlled by the PLC's. The initial configuration will include over 25 Allen-Bradley PLC-5 units. The PLCs in each subsystem are connected to personal computers (PC's) in the control room. The PC's provide graphical displays and operator interface. The Pc's are networked and share process data with each other and with a master control console and a large mimic panel

SCDAP/RELAP5/MOD3 code development and assessment

International Nuclear Information System (INIS)

Allison, C.M.; Heath, C.H.; Siefken, L.J.; Hohorst, J.K.

1991-01-01

The SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the Nuclear Regulatory Commission (NRC). SCDAP/RELAP5/MOD3, created in January, 1991, is the result of merging RELAP5/MOD3 with SCDAP and TRAP-MELT models from SCDAP/RELAP5/MOD2.5. The RELAP5 models calculate the overall RCS thermal-hydraulics, control system interactions, reactor kinetics, and the transport of noncondensible gases, fission products, and aerosols. The SCDAP models calculate the damage progression in the core structures, the formation, heatup, and melting of debris, and the creep rupture failure of the lower head and other RCS structures. The TRAP-MELT models calculate the deposition of fission products upon aerosols or structural surfaces; the formation, growth, or deposition of aerosols; and the evaporation of species from surfaces. The systematic assessment of modeling uncertainties in SCDAP/RELAP5 code is currently underway. This assessment includes (a) the evaluation of code-to-data comparisons using stand-alone SCDAP and SCDAP/RELAP5/MOD3, (b) the estimation of modeling and experimental uncertainties, and (c) the determination of the influence of those uncertainties on predicted severe accident behavior

Algebraic Structures on MOD Planes

OpenAIRE

Kandasamy, Vasantha; Ilanthenral, K.; Smarandache, Florentin

2015-01-01

Study of MOD planes happens to a very recent one. In this book, systematically algebraic structures on MOD planes like, MOD semigroups, MOD groups and MOD rings of different types are defined and studied. Such study is innovative for a large four quadrant planes are made into a small MOD planes. Several distinct features enjoyed by these MOD planes are defined, developed and described.

CORCON-MOD3: An integrated computer model for analysis of molten core-concrete interactions

International Nuclear Information System (INIS)

Bradley, D.R.; Gardner, D.R.; Brockmann, J.E.; Griffith, R.O.

1993-10-01

The CORCON-Mod3 computer code was developed to mechanistically model the important core-concrete interaction phenomena, including those phenomena relevant to the assessment of containment failure and radionuclide release. The code can be applied to a wide range of severe accident scenarios and reactor plants. The code represents the current state of the art for simulating core debris interactions with concrete. This document comprises the user's manual and gives a brief description of the models and the assumptions and limitations in the code. Also discussed are the input parameters and the code output. Two sample problems are also given

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

Vector models in RETRAN-02 MOD 2

International Nuclear Information System (INIS)

Kinnersly, S.R.

1985-06-01

The vector momentum model in RETRAN-02 allows momentum flux to be modelled in two dimensions. Vector models in RETRAN-2 are described, including both the actual implementation in the code and the specification given in the code manual. The vector momentum model is described in detail. Other models which use vector quantities include models for volume average flow, volume average slip velocity, volume average phase velocities and fill junction flows. Both code implementations and code manual descriptions are described and inconsistencies noted. The differences between the standard RETRA-02 Mod 2 version and the Winfrith version RETN2204 are noted. (U.K.)

NetMOD version 1.0 user's manual

Energy Technology Data Exchange (ETDEWEB)

Merchant, Bion John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-01-01

NetMOD (Network Monitoring for Optimal Detection) is a Java-based software package for conducting simulation of seismic networks. Specifically, NetMOD simulates the detection capabilities of seismic monitoring networks. Network simulations have long been used to study network resilience to station outages and to determine where additional stations are needed to reduce monitoring thresholds. NetMOD makes use of geophysical models to determine the source characteristics, signal attenuation along the path between the source and station, and the performance and noise properties of the station. These geophysical models are combined to simulate the relative amplitudes of signal and noise that are observed at each of the stations. From these signal-to-noise ratios (SNR), the probability of detection can be computed given a detection threshold. This manual describes how to configure and operate NetMOD to perform seismic detection simulations. In addition, NetMOD is distributed with a simulation dataset for the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO) International Monitoring System (IMS) seismic network for the purpose of demonstrating NetMOD's capabilities and providing user training. The tutorial sections of this manual use this dataset when describing how to perform the steps involved when running a simulation.

R-HyMOD: an R-package for the hydrological model HyMOD

Science.gov (United States)

Baratti, Emanuele; Montanari, Alberto

2015-04-01

A software code for the implementation of the HyMOD hydrological model [1] is presented. HyMOD is a conceptual lumped rainfall-runoff model that is based on the probability-distributed soil storage capacity principle introduced by R. J. Moore 1985 [2]. The general idea behind this model is to describe the spatial variability of some process parameters as, for instance, the soil structure or the water storage capacities, through probability distribution functions. In HyMOD, the rainfall-runoff process is represented through a nonlinear tank connected with three identical linear tanks in parallel representing the surface flow and a slow-flow tank representing groundwater flow. The model requires the optimization of five parameters: Cmax (the maximum storage capacity within the watershed), β (the degree of spatial variability of the soil moisture capacity within the watershed), α (a factor for partitioning the flow between two series of tanks) and the two residence time parameters of quick-flow and slow-flow tanks, kquick and kslow respectively. Given its relatively simplicity but robustness, the model is widely used in the literature. The input data consist of precipitation and potential evapotranspiration at the given time scale. The R-HyMOD package is composed by a 'canonical' R-function of HyMOD and a fast FORTRAN implementation. The first one can be easily modified and can be used, for instance, for educational purposes; the second part combines the R user friendly interface with a fast processing unit. [1] Boyle D.P. (2000), Multicriteria calibration of hydrological models, Ph.D. dissertation, Dep. of Hydrol. and Water Resour., Univ of Arizona, Tucson. [2] Moore, R.J., (1985), The probability-distributed principle and runoff production at point and basin scale, Hydrol. Sci. J., 30(2), 273-297.

Development and application of STCP Mod 1.1

International Nuclear Information System (INIS)

Hu Zhiyi

1990-01-01

A state-of-the-art Source Term Code Package-STCP Mod 1.1 is used for assessing the release of radioactive materials to the environment in severe reactor accidents. Its structure and function, and its installation and development on CYBER computer in China are introduced. Through calculations and analyses for 4 selected severe accident sequences from ZION Nuclear Power Plant of the U.S., it shows that the developed STCP Mod 1.1 is suitable to calculate varios scenarios as long as the selections of the model and parameters are reasonable

RELAP5/MOD3.2 investigation of loss of in-house supply power for WWER 1000/320V

International Nuclear Information System (INIS)

Gencheva, R.; Pavlova, M.; Groudev, P.

2001-01-01

This paper discusses the results of the thermal-hydraulic investigations of the 'Loss of in-house supply power' accident at the Kozloduy NPP Unit 6. The RELAP5/MOD3.2 computer code has been used to stimulate the loss of in-house supply power accident in a WWER 1000 Nuclear Power Plant model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for analyses of operational occurrences, abnormal events and design basis scenarios. It will provide a significant analytical capability for the Bulgarian technical specialists located at the Kozloduy NPP. The criteria used in selecting transient are: importance to safety, availability and suitability of data followed by suitability for RELAP5 code validation. The investigation of 'Loss of normal and reverse AC power' is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the WWER 1000 against experimental transient data obtained from Kozloduy NPP Unit 6. The comparisons between the RELAP5 results and the test data indicate good agreement

Assessment of PWR Steam Generator modelling in RELAP5/MOD2

International Nuclear Information System (INIS)

Putney, J.M.; Preece, R.J.

1993-06-01

An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3

International Code Assessment and Applications Program: Summary of code assessment studies concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B

International Nuclear Information System (INIS)

Schultz, R.R.

1993-12-01

Members of the International Code Assessment Program (ICAP) have assessed the US Nuclear Regulatory Commission (USNRC) advanced thermal-hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define user guidelines, and to determine the state of each code. The results of sixty-two code assessment reviews, conducted at INEL, are summarized. Code deficiencies are discussed and user recommended nodalizations investigated during the course of conducting the assessment studies and reviews are listed. All the work that is summarized was done using the RELAP5/MOD2, RELAP5/MOD3, and TRAC-B codes

Experiment data report for semiscale Mod-1 Test S-06-2 (LOFT counterpart test)

International Nuclear Information System (INIS)

Patton, M.L. Jr.; Collins, B.L.; Sackett, K.E.

1977-08-01

Recorded test data are presented for Test S-06-2 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying an hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-2 was conducted from initial conditions of 15 513 kPa and 563 K 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 broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 50% of the maximum peak power density

TRAC-PF1/MOD1 independent assessment: Semiscale Mod-2A intermediate break test S-IB-3

International Nuclear Information System (INIS)

Kmetyk, L.N.

1986-02-01

The TRAC-PF1/MOD1 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various system codes to predict the detailed thermal/hydraulic response of light water reactors during accident and off-normal conditions. The TRAC code is being assessed at SNLA against test data from various integral and separate effects test facilities. As part of this assessment matrix, an intermediate break test (S-IB-3), performed at the Semiscale Mod-2A facility, has been analyzed. Using an input model with a 3-D VESSEL component, the vessel and downcomer inventories during 3-IB-3 were generally well predicted, but the core heatup was underpredicted compared to data. An equivalent calculation with an all 1-D input model ran about twice as fast as our basecase analysis using a 3-D VESSEL in the input model, but the results of the two calculations diverged significantly for many parameters of interest, with the 3-D VESSEL model results in better agreement with data. 22 refs., 100 figs

NAUA-Mod 3 - A computer code for the description of the aerosol behaviour in a condensing atmosphere

International Nuclear Information System (INIS)

Bunz, H.; Koyro, M.; Schoeck, W.

1981-09-01

This report gives a description of the computer code NAUA-Mod 3. Its purpose is to calculate the behaviour of a polydisperse aerosol system in the containment of a light water reactor after a postulated core meltdown accident as a function of the time. The most important effect being added to those already taken into account in comparable computer codes is the steam condensation onto the particles. In the report the equations taken as basis of the code are given and the physical processes they are derived from are explained. Another main objekt of the report is the description of the numerical methods used as well as the input and output of the code. (orig.) [de

Relap5/mod2 post-test calculation of a loss of feedwater experiment at the Pactel test facility

Energy Technology Data Exchange (ETDEWEB)

Protze, M. [Siemens-KWU, Erlangen (Germany)

1995-12-31

Post-test calculations for verification purposes of the thermal hydraulic code RELAP5/MOD2 are of fundamental importance for the licensing procedure. The RELAP5/MOD2 code has a large international assessment base regarding western PWR. WWER-reactors are russian designed PWRs with some specific differences compared with the western PWR`s, especially the horizontal steam generators. For that reason some post-test calculations have to be performed to verify the RELAP5/MOD2 code for these WWER typical phenomena. The impact of the horizontal steam generators on the accident behaviour during transients or pipe ruptures on the secondary side is significant. The nodalization of the test facility PACTEL was chosen equally to WWER plant nodalization to verify the use of a coarse modelling of the steam generator secondary side for analyses of transient with decreasing water level in the SG secondary side. The calculational results showed a good compliance to the test results, demonstrating the correct use of a coarse nodalization. To sum up, the RELAP5/ MOD2 results met the test results appropriately thereby the RELAP5/ MOD2 code is validated for analyses of transients with decreasing water level in a horizontal steam generator secondary side. (orig.). 4 refs.

Relap5/mod2 post-test calculation of a loss of feedwater experiment at the Pactel test facility

Energy Technology Data Exchange (ETDEWEB)

Protze, M [Siemens-KWU, Erlangen (Germany)

1996-12-31

Post-test calculations for verification purposes of the thermal hydraulic code RELAP5/MOD2 are of fundamental importance for the licensing procedure. The RELAP5/MOD2 code has a large international assessment base regarding western PWR. WWER-reactors are russian designed PWRs with some specific differences compared with the western PWR`s, especially the horizontal steam generators. For that reason some post-test calculations have to be performed to verify the RELAP5/MOD2 code for these WWER typical phenomena. The impact of the horizontal steam generators on the accident behaviour during transients or pipe ruptures on the secondary side is significant. The nodalization of the test facility PACTEL was chosen equally to WWER plant nodalization to verify the use of a coarse modelling of the steam generator secondary side for analyses of transient with decreasing water level in the SG secondary side. The calculational results showed a good compliance to the test results, demonstrating the correct use of a coarse nodalization. To sum up, the RELAP5/ MOD2 results met the test results appropriately thereby the RELAP5/ MOD2 code is validated for analyses of transients with decreasing water level in a horizontal steam generator secondary side. (orig.). 4 refs.

Progress on MOD/RABiTS{sup TM} 2G HTS wire

Energy Technology Data Exchange (ETDEWEB)

Rupich, M.W.; Zhang, W.; Li, X.; Kodenkandath, T.; Verebelyi, D.T.; Schoop, U.; Thieme, C.; Teplitsky, M.; Lynch, J.; Nguyen, N.; Siegal, E.; Scudiere, J.; Maroni, V.; Venkataraman, K.; Miller, D.; Holesinger, T.G

2004-10-01

The development of the second generation (2G) high temperature superconducting wire has advanced beyond initial laboratory demonstrations and is now focused on developing and testing high critical current conductor designs required for commercial applications. The approach pursued at American Superconductor for 2G wire manufacturing is based on the combination of the RABiTS{sup TM} substrate-buffer technology with metal organic deposition (MOD) of the YBCO layer. This MOD/RABiTS{sup TM} approach has been demonstrated in 10 m lengths with critical currents of up to 184 A/cm-width ({approx}2.3 MA/cm{sup 2}) and in short length with critical currents of up to 270 A/cm-width ({approx}3.4 MA/cm{sup 2}). In addition to a high critical current, the superconducting wire must also meet stringent mechanical and electrical stability requirements that vary by application. Commercially viable architectures designed to meet these specifications have been fabricated and tested. Wires manufactured by this process have been successfully tested in prototype cable and coil applications.

Testing Gyrokinetics on C-Mod and NSTX

International Nuclear Information System (INIS)

Redi, M.H.; Dorland, W.; Fiore, C.L.; Stutman, D.; Baumgaertel, J.A.; Davis, B.; Kaye, S.M.; McCune, D.C.; Menard, J.; Rewoldt, G.

2005-01-01

Quantitative benchmarks of computational physics codes against experiment are essential for the credible application of such codes. Fluctuation measurements can provide necessary critical tests of nonlinear gyrokinetic simulations, but such require extraordinary computational resources. Linear micro-stability calculations with the GS2 [1] gyrokinetic code have been carried out for tokamak and ST experiments which exhibit internal transport barriers (ITB) and good plasma confinement. Qualitative correlation is found for improved confinement before and during ITB plasmas on Alcator C-Mod [2] and NSTX [3], with weaker long wavelength micro-instabilities in the plasma core regions. Mixing length transport models are discussed. The NSTX L-mode is found to be near marginal stability for kinetic ballooning modes. Fully electromagnetic, linear, gyrokinetic calculations of the Alcator C-Mod ITB during off-axis rf heating, following four plasma species and including the complete electron response show ITG/TEM microturbulence is suppressed in the plasma core and in the barrier region before barrier formation, without recourse to the usual requirements of velocity shear or reversed magnetic shear [4-5]. No strongly growing long or short wavelength drift modes are found in the plasma core but strong ITG/TEM and ETG drift wave turbulence is found outside the barrier region. Linear microstability analysis is qualitatively consistent with the experimental transport analysis, showing low transport inside and high transport outside the ITB region before barrier formation, without consideration of ExB shear stabilization

SCDAP/RELAP5/MOD3 code development

International Nuclear Information System (INIS)

Allison, C.M.; Siefken, J.L.; Coryell, E.W.

1992-01-01

The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC). Code development activities are currently focused on three main areas - (a) code usability, (b) early phase melt progression model improvements, and (c) advanced reactor thermal-hydraulic model extensions. This paper describes the first two activities. A companion paper describes the advanced reactor model improvements being performed under RELAP5/MOD3 funding

Analysis of steam generator loss-of-feedwater experiments with APROS and RELAP5/MOD3.1 computer codes

Energy Technology Data Exchange (ETDEWEB)

Virtanen, E.; Haapalehto, T. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Nuclear Energy, Lappeenranta (Finland)

1995-09-01

Three experiments were conducted to study the behavior of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes to that the results may be compared. Only the steam generator was modelled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments.

Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code

International Nuclear Information System (INIS)

Ramsthaler, J.A.; Lime, J.F.; Sahota, M.S.

1978-12-01

A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A

TRAC-PF1/MOD1 assessment at Los Alamos

International Nuclear Information System (INIS)

Knight, T.D.

1984-01-01

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in pressurized water reactors (PWRs). Over the past several years, four distinct versions of the code have been released; each new version introduced improvements to the existing models and numerics and added new models to extend the applications of the code. The first goal of the code was to analyze large-break loss-of-coolant accidents (LOCAs), and the TRAC-P1A and TRAC-PD2 codes primarily addressed the large-break LOCA. (The TRAC-PD2/MOD1 code is essentially the same as the TRAC-PD2 code but it also includes a released set of error corrections.) The TRAC-PF1 code contained major changes to the models and trips and to the numerical methods. These modifications enhanced the computational speed of the code and improved the application to small-break LOCAs. The TRAC-PF1/MOD1 code, the latest released version, added improved steam-generator modeling, a turbine component, and a control system together with modified constitutive relations to model the balance of plant on the secondary side and to extend the applications to non-LOCA transients. The TRAC-PF1/MOD1 code also contains reasonably general reactor-kinetics modeling to facilitate the simulation of transients with delayed scram or without scram. 13 references, 24 figures

Conceptual design of the 7 megawatt Mod-5B wind turbine generator

Science.gov (United States)

Douglas, R. R.

1982-01-01

Similar to MOD-2, the MOD-5B wind turbine generator system is designed for the sole purpose of providing electrical power for distribution by a major utility network. The objectives of the MOD-2 and MOD-5B programs are essentially identical with one important exception; the cost-of-electricity (COE) target is reduced from 4 cent/Kwhr on MOD-2 to 3 cent/Kwhr on MOD-5B, based on mid 1977 dollars and large quantity production. The MOD-5B concept studies and eventual concept selection confirmed that the program COE targets could not only be achieved but substantially bettered. Starting from the established MOD-2 technology as a base, this achievement resulted from a combination of concept changes, size changes, and design refinements. The result of this effort is a wind turbine system that can compete with conventional power generation over significant geographical areas, increasing commercial market potential by an order of magnitude.

Mod-5A wind turbine generator program design report. Volume 3: Final design and system description, book 2

Science.gov (United States)

1984-01-01

The design, development and analysis of the 7.3MW MOD-5A wind turbine generator is documented. The report is divided into four volumes: Volume 1 summarizes the entire MOD-5A program, Volume 2 discusses the conceptual and preliminary design phases, Volume 3 describes the final design of the MOD-5A, and Volume 4 contains the drawings and specifications developed for the final design. Volume 3, book 2 describes the performance and characteristics of the MOD-5A wind turbine generator in its final configuration. The subsystem for power generation, control, and instrumentation subsystems is described in detail. The manufacturing and construction plans, and the preparation of a potential site on Oahu, Hawaii, are documented. The quality assurance and safety plan, and analyses of failure modes and effects, and reliability, availability and maintainability are presented.

Experiment data report for semiscale Mod-1 test S-28-2 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Patton, M.L.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-2 of the Semiscale Mod-1 steam generator tube rupture 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-28-2 was conducted from initial conditions of 15 936 kPa and 558 K 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 broken loop cold leg 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. For Test S-28-2, accumulator injection into the intact loop hot leg was provided to simulate simulate the rupture of six steam generator tubes

Vertical downward subcooled bubbly flow modelling with RELAP5/MOD3.2.2 gamma

International Nuclear Information System (INIS)

Ristevski, R.; Parzer, I.; Markov, Z.

2000-01-01

The presented paper will consider the correlation for void fraction distribution in the subcooled boiling flow regime of downward liquid flow at low velocities. More specifically, it will focus on the choice of the most appropriate heat and mass transfer correlation. The experimental findings and theoretical consideration of these processes and phenomena will be compared with RELAP5/MOD3.2.2 Gamma predictions. (author)

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

Analysis of steam generator loss-of-feedwater experiments with APROS and RELAP5/MOD3.1 computer codes

International Nuclear Information System (INIS)

Virtanen, E.; Haapalehto, T.; Kouhia, J.

1997-01-01

Three experiments were conducted to study the behaviour of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes so that the results may be compared. Only the steam generator was modeled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments. (orig.)

Blow.MOD2: a program for blowdown transient calculations

International Nuclear Information System (INIS)

Doval, A.

1990-01-01

The BLOW.MOD2 program has been developed to calculate the blowdown phase in a pressurized vessel after a break/valve is opened. It is a one volume model where break height and flow area are specified. Moody critical flow model was adopted under saturation conditions for flow calculation through the break. Heat transfer from structures and internals have been taken into account. Long term depressurization results and a more complex model are compared satisfactorily. (Author)

RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

International Nuclear Information System (INIS)

Gencheva, Rositsa V.; Stefanova, Antoaneta E.; Groudev, Pavlin P.

2005-01-01

During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is 'Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed and Bleed procedure initiation. For the purpose of this, operator action with 'Reactor vessel off-gas valve - 0.032 m' opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety

Utilization of DRUFAN 01/MOD 02 computer code for the depressurization phase analysis of a postulated loss of coolant accident in Angra 2/3 Nuclear Power Plants

International Nuclear Information System (INIS)

Austregesilo Filho, H.; Figueiredo, P.J.M.

1985-08-01

The DRUFAN 01/Mod 2 developed by Gesellschaft fur Reaktorsicherheit (GRS) mbh to simulate thermohydraulic behavior of the primary circuit of PWR reactors, during the despressurization phase and initial refilling phase of loss of coolant accidents by great ruptures, is presented. The program simulates the system to be analysed by control volumes-concentrated parameters model - and it is based on numerical solution of conservation equations for mass of water, mass of vapor, quantities of motion and energy, and on the control volume homogeneity hypothesis. The possibilities of thermodynamic disequilibrium, determining mass transfer between liquid and vapor phases assuming that one saturated phase, are considered. The process of computer code implantation in the Honeywell Bull 64 DPS 7 system at CNEN, the modifications done into the program and the application to the despressurization phase analysis of a loss of coolant accident at Angra-2 and Angra-3 reactors are considered. (M.C.K.) [pt

Functional characterization of the Bradyrhizobium japonicum modA and modB genes involved in molybdenum transport.

Science.gov (United States)

Delgado, María J; Tresierra-Ayala, Alvaro; Talbi, Chouhra; Bedmar, Eulogio J

2006-01-01

A modABC gene cluster that encodes an ABC-type, high-affinity molybdate transporter from Bradyrhizobium japonicum has been isolated and characterized. B. japonicum modA and modB mutant strains were unable to grow aerobically or anaerobically with nitrate as nitrogen source or as respiratory substrate, respectively, and lacked nitrate reductase activity. The nitrogen-fixing ability of the mod mutants in symbiotic association with soybean plants grown in a Mo-deficient mineral solution was severely impaired. Addition of molybdate to the bacterial growth medium or to the plant mineral solution fully restored the wild-type phenotype. Because the amount of molybdate required for suppression of the mutant phenotype either under free-living or under symbiotic conditions was dependent on sulphate concentration, it is likely that a sulphate transporter is also involved in Mo uptake in B. japonicum. The promoter region of the modABC genes has been characterized by primer extension. Reverse transcription and expression of a transcriptional fusion, P(modA)-lacZ, was detected only in a B. japonicum modA mutant grown in a medium without molybdate supplementation. These findings indicate that transcription of the B. japonicum modABC genes is repressed by molybdate.

RELAP5/MOD2 code assessment using a LOFT L2-3 loss of coolant experiment

International Nuclear Information System (INIS)

Bang, Young Seok; Chung, Bub Dong; Kim, Hho Jung

1990-01-01

The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of the PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core in a reasonable range and that the code had deficiencies in the critical flow model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. A Sensitivity calculation with an updated version from RELAP5/MOD2 Cycle 36.04 improved the prediction of the rewet phenomena

Who owns the mods?

OpenAIRE

Kow, Yong Ming; Nardi, Bonnie

2010-01-01

Modding, the development of end user software extensions to commercial products, is popular among video gamers. Modders form communities to help each other. Mods can shape software products by weaving in contributions from users themselves based on their own experience of a product. The purpose of this paper is to investigate a conflict between a modding community and a gaming company which reveals contested issues of ownership and governance. We studied an online game, World of Warcraft, a l...

Assessment of RELAP5/MOD3.2 with condensation experiment in the presence of noncondensables in a vertical tube

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun Sik; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-12-31

The standard RELAP5/MOD3.2 code were assessed with the condensation experiment in the presence of noncondensable gas in a vertical tube of PCCS of CP-1300. There are two wall film condensation models, the default model and the alternative model, in RELAP5/MOD3.2. The experimental apparatus was modeled with the two models, and simulations were performed for several sub-tests to be compared with the experimental results. In overall sense the simulation results showed that the default model of RELAP5/MOD3.2 under-predicts the heat transfer coefficients, while the alternative model over-predicts them throughout the condensing tube. 10 refs., 6 figs. (Author)

Assessment of RELAP5/MOD3.2 with condensation experiment in the presence of noncondensables in a vertical tube

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun Sik; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

The standard RELAP5/MOD3.2 code were assessed with the condensation experiment in the presence of noncondensable gas in a vertical tube of PCCS of CP-1300. There are two wall film condensation models, the default model and the alternative model, in RELAP5/MOD3.2. The experimental apparatus was modeled with the two models, and simulations were performed for several sub-tests to be compared with the experimental results. In overall sense the simulation results showed that the default model of RELAP5/MOD3.2 under-predicts the heat transfer coefficients, while the alternative model over-predicts them throughout the condensing tube. 10 refs., 6 figs. (Author)

Evaluation of the applicability of cladding deformation model in RELAP5/MOD3.2 code for VVER-1000 fuel

International Nuclear Information System (INIS)

Vorob'ev, Yu.; Zhabin, O.

2015-01-01

Applicability of cladding deformation model in RELAP5/MOD3.2 code is analyzed for VVER-1000 fuel cladding from Zr+1%Nb alloy. Experimental data and calculation model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the cladding temperature range from 600 to 1200 deg C and pressure range from 1 to 12 MPa. Evaluation results demonstrate limited applicability of built-in RELAP5/MOD3.2 cladding deformation model to the estimation of Zr+1%Nb cladding rupture conditions. The limitations found shall be considered in application of RELAP5/MOD3.2 cladding deformation model in the design-basis accident analysis of VVER reactors

Alcator C-Mod Tokamak

Data.gov (United States)

Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...

Core heatup prediction during SB LOCA with RELAP5/MOD3.2.2 Gamma

International Nuclear Information System (INIS)

Parzer, I.; Mavko, B.; Petelin, S.

2001-01-01

The paper focuses on the phenomena leading to core uncovering and heatup during the SB LOCA and the ability of RELAP5/MOD3.2.2 Gamma to predict core overheating. The code prediction has been compared to the three experiments, one conducted on the separate effect test facility NEPTUN in Switzerland and the other two conducted on two integral test facilities, PMK-2 in Hungary and PACTEL facility in Finland. In the case of a series of boiloff experiments performed on the NEPTUN test facility the influence of the two correlations available in MOD3.2.2 Gamma for determining interphase drag has been studied. In the case of IAEA-SPE-4 experiment simulation on PMK-2 facility the main goal of the analysis was to study the adequate modeling of the hexagonal core channel with 19-rod bundle and the phenomena during the core uncovering. The third analyzed experiment, OECD-ISP-33, was performed on PACTEL facility to study different natural circulation modes during SB LOCA. The analysis also focused on the final stage of this SB LOCA experiment, when core dryout and heatup was observed due to gradual emptying of the primary system. Following the experience the appropriate modeling options have been used to achieve better representation of the important phenomena during the SB LOCA.(author)

An experimental assessment of methods used to compute secondary electron emission yield for tungsten and molybdenum electrodes based on exposure to Alcator C-Mod scrape-off layer plasmas

Science.gov (United States)

McCarthy, W.; LaBombard, B.; Brunner, D.; Kuang, A. Q.

2018-03-01

Plasma potentials computed from Langmuir probe data rely on a method to account for secondary electron emission (SEE) from the electrodes. However, significant variations exist among published models for SEE and the reported experimental parameters used to evaluate them. As a means to critically assess SEE computation methods, two of four tungsten electrodes on a Langmuir-Mach probe head were replaced with molybdenum and exposed to Alcator C-Mod boundary plasmas where electron temperatures exceed 50 eV and SEE becomes significant. In this situation, plasma potentials computed for either material should be identical—the SEE evaluation method should properly account for the differences in SEE yields. Of the six methods used to compute SEE, two are found to produce consistent results (Sternglass model with Bronstein experimental parameters and Young-Dekker model with Bronstein experimental parameters). In contrast, the method previously used for C-Mod data analysis (Sternglass model with Kollath parameters) was found to be inconsistent. We have since adopted Young-Dekker-Bronstein as the preferred method.

RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

Energy Technology Data Exchange (ETDEWEB)

Bayless, P.D.

2003-01-17

The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.

RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

International Nuclear Information System (INIS)

Bayless, P.D.

2003-01-01

The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior

Radon transport modelling: User's guide to RnMod3d

DEFF Research Database (Denmark)

Andersen, Claus Erik

2000-01-01

RnMod3d is a numerical computer model of soil-gas and radon transport in porous media. It can be used, for example, to study radon entry from soil into houses in response to indoor-outdoor pressure differences or changes in atmospheric pressure. It canalso be used for flux calculations of radon...... decay, diffusion and advection of radon can be solved. Moisture is included in the model, and partitioning ofradon between air, water and soil grains (adsorption) is taken into account. Most parameters can change in time and space, and transport parameters (diffusivity and permeability) may...... be anisotropic. This guide includes benchmark tests based on simpleproblems with known solutions. RnMod3d has also been part of an international model intercomparison exercise based on more complicated problems without known solutions. All tests show that RnMod3d gives results of good quality....

Cloud-based uniform ChIP-Seq processing tools for modENCODE and ENCODE.

Science.gov (United States)

Trinh, Quang M; Jen, Fei-Yang Arthur; Zhou, Ziru; Chu, Kar Ming; Perry, Marc D; Kephart, Ellen T; Contrino, Sergio; Ruzanov, Peter; Stein, Lincoln D

2013-07-22

Funded by the National Institutes of Health (NIH), the aim of the Model Organism ENCyclopedia of DNA Elements (modENCODE) project is to provide the biological research community with a comprehensive encyclopedia of functional genomic elements for both model organisms C. elegans (worm) and D. melanogaster (fly). With a total size of just under 10 terabytes of data collected and released to the public, one of the challenges faced by researchers is to extract biologically meaningful knowledge from this large data set. While the basic quality control, pre-processing, and analysis of the data has already been performed by members of the modENCODE consortium, many researchers will wish to reinterpret the data set using modifications and enhancements of the original protocols, or combine modENCODE data with other data sets. Unfortunately this can be a time consuming and logistically challenging proposition. In recognition of this challenge, the modENCODE DCC has released uniform computing resources for analyzing modENCODE data on Galaxy (https://github.com/modENCODE-DCC/Galaxy), on the public Amazon Cloud (http://aws.amazon.com), and on the private Bionimbus Cloud for genomic research (http://www.bionimbus.org). In particular, we have released Galaxy workflows for interpreting ChIP-seq data which use the same quality control (QC) and peak calling standards adopted by the modENCODE and ENCODE communities. For convenience of use, we have created Amazon and Bionimbus Cloud machine images containing Galaxy along with all the modENCODE data, software and other dependencies. Using these resources provides a framework for running consistent and reproducible analyses on modENCODE data, ultimately allowing researchers to use more of their time using modENCODE data, and less time moving it around.

Apo and ligand-bound structures of ModA from the archaeon Methanosarcina acetivorans.

Science.gov (United States)

Chan, Sum; Giuroiu, Iulia; Chernishof, Irina; Sawaya, Michael R; Chiang, Janet; Gunsalus, Robert P; Arbing, Mark A; Perry, L Jeanne

2010-03-01

The trace-element oxyanion molybdate, which is required for the growth of many bacterial and archaeal species, is transported into the cell by an ATP-binding cassette (ABC) transporter superfamily uptake system called ModABC. ModABC consists of the ModA periplasmic solute-binding protein, the integral membrane-transport protein ModB and the ATP-binding and hydrolysis cassette protein ModC. In this study, X-ray crystal structures of ModA from the archaeon Methanosarcina acetivorans (MaModA) have been determined in the apoprotein conformation at 1.95 and 1.69 A resolution and in the molybdate-bound conformation at 2.25 and 2.45 A resolution. The overall domain structure of MaModA is similar to other ModA proteins in that it has a bilobal structure in which two mixed alpha/beta domains are linked by a hinge region. The apo MaModA is the first unliganded archaeal ModA structure to be determined: it exhibits a deep cleft between the two domains and confirms that upon binding ligand one domain is rotated towards the other by a hinge-bending motion, which is consistent with the 'Venus flytrap' model seen for bacterial-type periplasmic binding proteins. In contrast to the bacterial ModA structures, which have tetrahedral coordination of their metal substrates, molybdate-bound MaModA employs octahedral coordination of its substrate like other archaeal ModA proteins.

RELAP5/MOD2 post-test calculation of the OECD LOFT experiment LP-SB-2

International Nuclear Information System (INIS)

Perez, J.; Mendizabal, R.

1992-04-01

This document presents the analysis of the OECD LOFT LP-SB-2 Experiment performed by the Consejo de Seguridad Nuclear of Spain working group making use of RELAP5/MOD2 in the frame of the Spanish LOFT Project. LB-SB-2 experiment studies the effect of a delayed pump trip in a small break LOCA scenario with a 3-inch equivalent diameter break in the hot leg of a commercial PWR

Apo and ligand-bound structures of ModA from the archaeon Methanosarcina acetivorans

International Nuclear Information System (INIS)

Chan, Sum; Giuroiu, Iulia; Chernishof, Irina; Sawaya, Michael R.; Chiang, Janet; Gunsalus, Robert P.; Arbing, Mark A.; Perry, L. Jeanne

2010-01-01

Crystal structures of ModA from M. acetivorans in the apo and ligand-bound conformations confirm domain rotation upon ligand binding. The trace-element oxyanion molybdate, which is required for the growth of many bacterial and archaeal species, is transported into the cell by an ATP-binding cassette (ABC) transporter superfamily uptake system called ModABC. ModABC consists of the ModA periplasmic solute-binding protein, the integral membrane-transport protein ModB and the ATP-binding and hydrolysis cassette protein ModC. In this study, X-ray crystal structures of ModA from the archaeon Methanosarcina acetivorans (MaModA) have been determined in the apoprotein conformation at 1.95 and 1.69 Å resolution and in the molybdate-bound conformation at 2.25 and 2.45 Å resolution. The overall domain structure of MaModA is similar to other ModA proteins in that it has a bilobal structure in which two mixed α/β domains are linked by a hinge region. The apo MaModA is the first unliganded archaeal ModA structure to be determined: it exhibits a deep cleft between the two domains and confirms that upon binding ligand one domain is rotated towards the other by a hinge-bending motion, which is consistent with the ‘Venus flytrap’ model seen for bacterial-type periplasmic binding proteins. In contrast to the bacterial ModA structures, which have tetrahedral coordination of their metal substrates, molybdate-bound MaModA employs octahedral coordination of its substrate like other archaeal ModA proteins

Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

Energy Technology Data Exchange (ETDEWEB)

Wissinger, G.; Klingenfus, J. [B & W Nuclear Technologies, Lynchburg, VA (United States)

1995-09-01

B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

Prediction of thermal-Hydraulic phenomena in the LBLOCA experiment L2-3 using RELAP5/MOD2

International Nuclear Information System (INIS)

Bang, Young Seok; Chung, Bub Dong; Kim, Hho Jung

1991-01-01

The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of a PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core reasonably and that the code had deficiencies in the critical flow model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. Improvement of the code, RELAP5/MOD2 Cycle 36.04, based on the sensitivity study increased the accuracy of the prediction of the rewet phenomena. (Author)

Developmental assessment of RELAP5/MOD3 using the semiscale natural circulation tests

International Nuclear Information System (INIS)

Carlson, K.E.

1990-01-01

A code development effort creating RELAP5/MOD3 from RELAP5/MOD2 has been completed. Upon completion, a developmental assessment task was performed. One of the problems used for the developmental assessment was the Semiscale Natural Circulation Test. Calculated results from RELAP5/MOD3 are compared to measured data and previously calculated results from RELAP5/MOD2. 10 refs., 6 figs., 1 tab

RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

International Nuclear Information System (INIS)

1976-06-01

A discussion is presented of the use of the RELAP4/MOD5 computer program in simulating the thermal-hydraulic behavior of light-water reactor systems when subjected to postulated transients such as a LOCA, pump failure, or nuclear excursion. The volume is divided into main sections which cover: (1) program description, (2) input data, (3) problem initialization, (4) user guidelines, (5) output discussion, (6) source program description, (7) implementation requirements, (8) data files, (9) description of PLOTR4M, (10) description of STH20, (11) summary flowchart, (12) sample problems, (13) problem definition, and (14) problem input

Molybdate binding by ModA, the periplasmic component of the Escherichia coli mod molybdate transport system.

Science.gov (United States)

Imperial, J; Hadi, M; Amy, N K

1998-03-13

ModA, the periplasmic-binding protein of the Escherichia coli mod transport system was overexpressed and purified. Binding of molybdate and tungstate to ModA was found to modify the UV absorption and fluorescence emission spectra of the protein. Titration of these changes showed that ModA binds molybdate and tungstate in a 1:1 molar ratio. ModA showed an intrinsic fluorescence emission spectrum attributable to its three tryptophanyl residues. Molybdate binding caused a conformational change in the protein characterized by: (i) a shift of tryptophanyl groups to a more hydrophobic environment; (ii) a quenching (at pH 5.0) or enhancement (at pH 7.8) of fluorescence; and (iii) a higher availability of tryptophanyl groups to the polar quencher acrylamide. The tight binding of molybdate did not allow an accurate estimation of the binding constants by these indirect methods. An isotopic binding method with 99MoO42- was used for accurate determination of KD (20 nM) and stoichiometry (1:1 molar ratio). ModA bound tungstate with approximately the same affinity, but did not bind sulfate or phosphate. These KDs are 150- to 250-fold lower than those previously reported, and compatible with the high molybdate transport affinity of the mod system. The affinity of ModA for molybdate was also determined in vivo and found to be similar to that determined in vitro. Copyright 1998 Elsevier Science B.V.

RELOS.MOD2: a code system for the determination of instationary fission product releases from molten pools

International Nuclear Information System (INIS)

Kortz, Ch.; Koch, M.K.; Unger, H.; Funke, F.

1999-01-01

For the assessment of molten corium pool source terms, a mechanistic model has been developed to describe the transport of fission products from liquid corium pool surfaces into a colder gas atmosphere. Modelling is based on an approach for diffusive and convective transport processes coupled with thermochemical equilibrium considerations enabling detailed speciation analyses of the fission products released. Both have been implemented into the code system RELOS.MOD2. RELOS.MOD2 sensitivity calculations on possible effects of anticipated uncertainties in the thermo-chemical data on the fission product release predictions are presented. (author)

MOD silver metallization for photovoltaics

Science.gov (United States)

Vest, G. M.; Vest, R. W.

1984-01-01

The development of flat plate solar arrays is reported. Photovoltaic cells require back side metallization and a collector grid system on the front surface. Metallo-organic decomposition (MOD) silver films can eliminate most of the present problems with silver conductors. The objectives are to: (1) identify and characterize suitable MO compounds; (2) develop generic synthesis procedures for the MO compounds; (3) develop generic fabrication procedures to screen printable MOD silver inks; (4) optimize processing conditions to produce grid patterns and photovoltaic cells; and (5) develop a model which describes the adhesion between the fired silver film and the silicon surface.

Comparison of SCDAP/RELAP5/MOD3 to TRAC-PF1/MOD1 for timing analysis of PWR fuel pin failures

International Nuclear Information System (INIS)

Jones, K.R.; Katsma, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.

1991-01-01

A comparison has been made of SCDAP/RELAP5/MOD3- and TRAC-PF1/MOD1- based calculations of the fuel pin failure timing (time from containment isolation signal to first fuel pin failure) in a loss-of-coolant accident (LOCA). The two codes were used to calculate the thermal-hydraulic boundary conditions for a complete, double-ended, offset-shear break of a cold leg in a Westinghouse 4-loop pressurized water reactor. Both calculations used the FRAPCON-2 code to calculate the steady-state fuel rod behavior and the FRAP-T6 code to calculate the transient fuel rod behavior. The analysis was performed for 16 combinations of fuel burnups and power peaking factors extending up to the Technical Specifications limits. While all calculations were made on a best-estimate basis, the SCDAP/RELAP5/MOD3 code has not yet been fully assessed for large-break LOCA analysis. The results indicate that SCDAP/RELAP5/MOD3 yields conservative fuel pin failure timing results in comparison to those generated using TRAC-PF1/MOD1. 7 refs., 5 figs

Conceptual design of the 6 MW Mod-5A wind turbine generator

Science.gov (United States)

Barton, R. S.; Lucas, W. C.

1982-01-01

The General Electric Company, Advanced Energy Programs Department, is designing under DOE/NASA sponsorship the MOD-5A wind turbine system which must generate electricity for 3.75 cent/KWH (1980) or less. During the Conceptual Design Phase, completed in March, 1981, the MOD-5A WTG system size and features were established as a result of tradeoff and optimization studies driven by minimizing the system cost of energy (COE). This led to a 400' rotor diameter size. The MOD-5A system which resulted is defined in this paper along with the operational and environmental factors that drive various portions of the design. Development of weight and cost estimating relationships (WCER's) and their use in optimizing the MOD-5A are discussed. The results of major tradeoff studies are also presented. Subsystem COE contributions for the 100th unit are shown along with the method of computation. Detailed descriptions of the major subsystems are given, in order that the results of the various trade and optimization studies can be more readily visualized.

Extension of the lod score: the mod score.

Science.gov (United States)

Clerget-Darpoux, F

2001-01-01

In 1955 Morton proposed the lod score method both for testing linkage between loci and for estimating the recombination fraction between them. If a disease is controlled by a gene at one of these loci, the lod score computation requires the prior specification of an underlying model that assigns the probabilities of genotypes from the observed phenotypes. To address the case of linkage studies for diseases with unknown mode of inheritance, we suggested (Clerget-Darpoux et al., 1986) extending the lod score function to a so-called mod score function. In this function, the variables are both the recombination fraction and the disease model parameters. Maximizing the mod score function over all these parameters amounts to maximizing the probability of marker data conditional on the disease status. Under the absence of linkage, the mod score conforms to a chi-square distribution, with extra degrees of freedom in comparison to the lod score function (MacLean et al., 1993). The mod score is asymptotically maximum for the true disease model (Clerget-Darpoux and Bonaïti-Pellié, 1992; Hodge and Elston, 1994). Consequently, the power to detect linkage through mod score will be highest when the space of models where the maximization is performed includes the true model. On the other hand, one must avoid overparametrization of the model space. For example, when the approach is applied to affected sibpairs, only two constrained disease model parameters should be used (Knapp et al., 1994) for the mod score maximization. It is also important to emphasize the existence of a strong correlation between the disease gene location and the disease model. Consequently, there is poor resolution of the location of the susceptibility locus when the disease model at this locus is unknown. Of course, this is true regardless of the statistics used. The mod score may also be applied in a candidate gene strategy to model the potential effect of this gene in the disease. Since, however, it

RELAP-5/MOD 3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

Energy Technology Data Exchange (ETDEWEB)

Paul D. Bayless

2003-01-01

The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.

Mod-5A wind turbine generator program design report. Volume 2: Conceptual and preliminary design, book 2

Science.gov (United States)

1984-01-01

The design, development and analysis of the 7.3 MW MOD-5A wind tunnel generator is documented. There are four volumes. In Volume 2, book 2 the requirements and criteria for the design are presented. The development tests, which determined or characterized many of the materials and components of the wind turbine generator, are described.

Expression, purification and DNA-binding activities of two putative ModE proteins of Herbaspirillum seropedicae (Burkholderiales, Oxalobacteraceae

Directory of Open Access Journals (Sweden)

André L.F. Souza

2008-01-01

Full Text Available In prokaryotes molybdenum is taken up by a high-affinity ABC-type transporter system encoded by the modABC genes. The endophyte β-Proteobacterium Herbaspirillum seropedicae has two modABC gene clusters and two genes encoding putative Mo-dependent regulator proteins (ModE1 and ModE2. Analysis of the amino acid sequence of the ModE1 protein of H. seropedicae revealed the presence of an N-terminal domain containing a DNA-binding helix-turn-helix motif (HTH and a C-terminal domain with a molybdate-binding motif. The second putative regulator protein, ModE2, contains only the helix-turn-helix motif, similar to that observed in some sequenced genomes. We cloned the modE1 (810 bp and modE2 (372 bp genes and expressed them in Escherichia coli as His-tagged fusion proteins, which we subsequently purified. The over-expressed recombinant His-ModE1 was insoluble and was purified after solubilization with urea and then on-column refolded during affinity chromatography. The His-ModE2 was expressed as a soluble protein and purified by affinity chromatography. These purified proteins were analyzed by DNA band-shift assays using the modA2 promoter region as probe. Our results indicate that His-ModE1 and His-ModE2 are able to bind to the modA2 promoter region, suggesting that both proteins may play a role in the regulation of molybdenum uptake and metabolism in H. seropedicae.

Overexpression, purification, and partial characterization of ADP-ribosyltransferases modA and modB of bacteriophage T4.

Science.gov (United States)

Tiemann, B; Depping, R; Rüger, W

1999-01-01

There is increasing experimental evidence that ADP-ribosylation of host proteins is an important means to regulate gene expression of bacteriophage T4. Surprisingly, this phage codes for three different ADP-ribosyltransferases, gene products Alt, ModA, and ModB, modifying partially overlapping sets of host proteins. While gene product Alt already has been isolated as a recombinant protein and its action on host RNA polymerases and transcription regulation have been studied, the nucleotide sequences of the two mod genes was published only recently. Their mode of action in the course of the infection cycle and the consequences of the ADP-ribosylations catalyzed by these enzymes remain to be investigated. Here we describe the cloning of the genes, the overexpression, purification, and partial characterization of ADP-ribosyltransferases ModA and ModB. Both proteins seem to act independently, and the ADP-ribosyl moieties are transferred to different sets of host proteins. While gene product ModA, similarly to the Alt protein, acts also on the alpha-subunit of host RNA polymerase, the ModB activity serves another set of proteins, one of which was identified as the S1 protein associated with the 30S subunit of the E. coli ribosomes.

TMI-2 analysis using SCDAP/RELAP5/MOD3.1

International Nuclear Information System (INIS)

Hohorst, J.K.; Polkinghorne, S.T.; Siefken, L.J.; Allison, C.M.; Dobbe, C.A.

1994-11-01

SCDAP/RELAP5/MOD3.1, an integrated thermal hydraulic analysis code developed primarily to simulate severe accidents in nuclear power plants, was used to predict the progression of core damage during the TMI-2 accident. The version of the code used for the TMI-2 analysis described in this paper includes models to predict core heatup, core geometry changes, and the relocation of molten core debris to the lower plenum of the reactor vessel. This paper describes the TMI-2 input model, initial conditions, boundary conditions, and the results from the best-estimate simulation of Phases 1 to 4 of the TMI-2 accident as well as the results from several sensitivity calculations

RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

International Nuclear Information System (INIS)

1976-09-01

This portion of the RELAP4/MOD5 User's Manual presents the details of setting up and entering the reactor model to be evaluated. The input card format and arrangement is presented in depth, including not only cards for data but also those for editing and restarting. Problem initalization including pressure distribution and energy balance is discussed. A section entitled ''User Guidelines'' is included to provide modeling recommendations, analysis and verification techniques, and computational difficulty resolution. The section is concluded with a discussion of the computer output form and format

Mod 1 ICS TI Report: ICS Conversion of a 140% HPGe Detector

Energy Technology Data Exchange (ETDEWEB)

Bounds, John Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-07-05

This report evaluates the Mod 1 ICS, an electrically cooled 140% HPGe detector. It is a custom version of the ORTEC Integrated Cooling System (ICS) modified to make it more practical for us to use in the field. Performance and operating characteristics of the Mod 1 ICS are documented, noting both pros and cons. The Mod 1 ICS is deemed a success. Recommendations for a Mod 2 ICS, a true field prototype, are provided.

Experimental and computational evaluation of neutrals in the Alcator C-Mod edge pedestal

Science.gov (United States)

Hughes, J. W.; Mossessian, D.; Labombard, B.; Terry, J.

2004-11-01

Pedestal-forming edge transport barriers (ETBs) in tokamak plasmas and the physics governing them are linked to the enhancement of confinement obtained in H-mode plasmas. Studies on Alcator C-Mod employ experimental measurements and simple 1-D transport models in order to better understand ETB physics. We examine the influences of ionization and charge exchange on the pedestals in electron density and temperature. Routine measurements from edge Thomson scattering (ETS) give pedestal scalings with global plasma parameters, while individual ETS profiles are combined with scanning Langmuir probe data and optical D_α emissivity measurements to give atomic density profiles and the associated radial distribution of the ionization source rate. From H-mode profiles of these quantities a well in effective plasma diffusivity is calculated, and is shown to systematically vary as the confinement regime is varied from ELM-free to EDA. Experimental work is supplemented with modeling and computation of edge neutral transport via KN1D, a kinetic solver for atomic and molecular distribution functions in slab geometry. The level of agreement between experiment and model is encouraging.

Adjoint sensitivity analysis of the RELAPS/MOD3.2 two-fluid thermal-hydraulic code system

International Nuclear Information System (INIS)

Ionescu-Bujor, M.

2000-10-01

This work presents the implementation of the Adjoint Sensitivity Analysis Procedure (ASAP) for the non-equilibrium, non-homogeneous two-fluid model, including boron concentration and non-condensable gases, of the RELAP5/MOD3.2 code. The end-product of this implementation is the Adjoint Sensitivity Model (ASM-REL/TF), which is derived for both the differential and discretized equations underlying the two-fluid model with non-condensable(s). The consistency requirements between these two representations are also highlighted. The validation of the ASM-REL/TF has been carried out by using sample problems involving: (i) liquid-phase only, (ii) gas-phase only, and (iii) two-phase mixture (of water and steam). Thus the 'Two-Loops with Pumps' sample problem supplied with RELAP5/MOD3.2 has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when only the liquid-phase is present. Furthermore, the 'Edwards Pipe' sample problem, also supplied with RELAP5/MOD3.2, has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when both (i.e., liquid and gas) phases are present. In addition, the accuracy and stability of the numerical solution of the ASM-REL/TF have been verified when only the gas-phase is present by using modified 'Two-Loops with Pumps' and the 'Edwards Pipe' sample problems in which the liquid and two-phase fluids, respectively, were replaced by pure steam. The results obtained for these sample problems depict typical sensitivities of junction velocities and volume-averaged pressures to perturbations in initial conditions, and indicate that the numerical solution of the ASM-REL/TF is as robust, stable, and accurate as the original RELAP5/MOD3.2 calculations. In addition, the solution of the ASM-REL/TF has been used to calculate sample sensitivities of volume-averaged pressures to variations in the pump head. (orig.) [de

Nonequilibrium constitutive models for RELAP5/MOD2

International Nuclear Information System (INIS)

Lin, J.C.; Trapp, J.A.; Riemke, R.A.; Ransom, V.H.

1983-01-01

RELAP5/MOD2 is a new version of RELAP5 containing improved modeling features that provide a generic pressurized-water transient simulation capability. In particular, the nonequilibrium modeling capability has been generalized to include conditions that occur in operational transients including repressurization and emergency-feed-water injection with loss-of-coolant accidents. The improvements include addition of a second energy equation to the hydrodynamic model, addition of nonequilibrium heat-transfer models, and the associated nonequilibrium vapor-generation models. The objective of this paper is to describe these models and to report the developmental assessment results obtained from similar of several separate effects experiments. The assessment shows that RELAP5 calculated results are in good agreement with data and the nonequilibrium phenomena are properly modeled

The WECHSL-Mod2 code: A computer program for the interaction of a core melt with concrete including the long term behavior

International Nuclear Information System (INIS)

Reimann, M.; Stiefel, S.

1989-06-01

The WECHSL-Mod2 code is a mechanistic computer code developed for the analysis of the thermal and chemical interaction of initially molten LWR reactor materials with concrete in a two-dimensional, axisymmetrical concrete cavity. The code performs calculations from the time of initial contact of a hot molten pool over start of solidification processes until long term basemat erosion over several days with the possibility of basemat penetration. The code assumes that the metallic phases of the melt pool form a layer at the bottom overlayed by the oxide melt atop. Heat generation in the melt is by decay heat and chemical reactions from metal oxidation. Energy is lost to the melting concrete and to the upper containment by radiation or evaporation of sumpwater possibly flooding the surface of the melt. Thermodynamic and transport properties as well as criteria for heat transfer and solidification processes are internally calculated for each time step. Heat transfer is modelled taking into account the high gas flux from the decomposing concrete and the heat conduction in the crusts possibly forming in the long term at the melt/concrete interface. The WECHSL code in its present version was validated by the BETA experiments. The test samples include a typical BETA post test calculation and a WECHSL application to a reactor accident. (orig.) [de

Simulation of the postulated stopping accident of the bombs of the primary circuit of Angra 2 with the code RELAP5/MOD3.2

International Nuclear Information System (INIS)

Andrade, Delvonei A.; Sabundjian, Gaiane; Madeira, Alzira A.; Pereira, Luiz Carlos M.; Borges, Ronaldo C.; Lapa, Nelbia S.

2001-01-01

This work presents the simulation of an anticipated transient for Angra 2 Nuclear Power Plant, where the coast down of the four reactor coolant pumps is verified. The best estimate thermal hydraulic system code RELAP5/MOD3.2 was used on this frame. A multi-purpose nodalization of Angra 2 was developed to simulate a comprehensive set of operational transients and accidents with RELAP5/MOD3.2 code. The overall objective of this work is to provide independent accident evaluation and further operational behavior follow-up to support the licensing process of the plant. (author)

RETRAN-3D MOD003 Peach Bottom Turbine Trip 2 Multidimensional Kinetics Analysis Models and Results

International Nuclear Information System (INIS)

Mori, Michitsugu; Ogura, Katsunori; Gose, Garry C.; Wu, J.-Y.

2003-01-01

An analysis of the Peach Bottom Unit 2 Turbine Trip Test 2 (PB2/TT2) has been performed using RETRAN-3D MOD003. The purpose of the analysis was to investigate the PB2/TT2 overpressurization transient using the RETRAN-3D multidimensional kinetics model

Experiment data report for semiscale Mod-2A primary feed and bleed experiment series (Tests S-SR-1 and S-SR-2)

International Nuclear Information System (INIS)

Fogdall, S.P.

1982-10-01

This report presents test data recorded for Tests S-SR-1 and S-SR-2 of the Semiscale Mod-2A Primary Feed and Bleed Tests. These tests are part of a series of Semiscale tests that investigate the thermal-hydraulic phenomena resulting from a hypothesized loss-of-coolant accident (LOCA) or abnormal operating transient. These tests provide experimental data for assessing the analytical capability of computer codes used in LOCA and operational transient analysis. The primary objectives of Tests S-SR-1 and -2 were to provide data on primary system recovery through the use of primary feed and bleed cooling, with no heat transfer to the secondaries. Data was obtained using high- and low-head pump curves for the safety injection (SI) pumps. This report presents the uninterpreted data from Tests S-SR-1 and -2 for analysis. The data, presented as graphs in engineering units, have been analyzed only to the extent necessary to ensure that they are reasonable and consistent

TRAC-PF1/MOD3 calculations of Savannah River Laboratory Rig FA single-annulus heated experiments

International Nuclear Information System (INIS)

Fischer, S.R.; McDaniel, C.K.

1992-01-01

This paper presents the results of TRAC-PF1/MOD3 benchmarks of the Rig FA experiments performed at the Savannah River Laboratory to simulate prototypic reactor fuel assembly behavior over a range of fluid conditions typical of the emergency cooling system (ECS) phase of a loss-of-coolant accident (LOCA). The primary purpose of this work was to use the SRL Rig FA tests to qualify the TRAC-PF1/MOD3 computer code and models for computing Mark-22 fuel assembly LOCA/ECS power limits. This qualification effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to independently confirm power limits for the Savannah River Site K Reactor. The results of this benchmark effort as discussed in this paper demonstrate that TRAC/PF1/MOD3 coupled with proper modeling is capable of simulating thermal-hydraulic phenomena typical of that encountered in Mark-22 fuel assembly during LOCA/ECS conditions

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

Stability of Microturbulent Drift Modes during Internal Transport Barrier Formation in the Alcator C-Mod Radio Frequency Heated H-mode

International Nuclear Information System (INIS)

Redi, M.H.; Fiore, C.L.; Dorland, W.; Mikkelsen, D.R.; Rewoldt, G.; Bonoli, P.T.; Ernst, D.R.; Rice, J.E.; Wukitch, S.J.

2003-01-01

Recent H-mode experiments on Alcator C-Mod [I.H. Hutchinson, et al., Phys. Plasmas 1 (1994) 1511] which exhibit an internal transport barrier (ITB), have been examined with flux tube geometry gyrokinetic simulations, using the massively parallel code GS2 [M. Kotschenreuther, G. Rewoldt, and W.M. Tang, Comput. Phys. Commun. 88 (1995) 128]. The simulations support the picture of ion/electron temperature gradient (ITG/ETG) microturbulence driving high xi/ xe and that suppressed ITG causes reduced particle transport and improved ci on C-Mod. Nonlinear calculations for C-Mod confirm initial linear simulations, which predicted ITG stability in the barrier region just before ITB formation, without invoking E x B shear suppression of turbulence. Nonlinear fluxes are compared to experiment, which both show low heat transport in the ITB and higher transport within and outside of the barrier region

Assessment of RELAP5/MOD3 with condensation experiment for pure steam condensation in a vercal tube

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Jae; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-12-31

The film condensation models in RELAP5/MOD3.1 and RELAP5/MOD3.2 are assessed with the data of experiment performed in the scaled down condensation experimental facility with a single vertical tube of inner diameter of 46 mm in the range of pressure 0.1 {approx} 7.5 MPa for the PSCS(Passive Secondary Condenser System). Both MOD3.1 and MOD3.2 don`t shows any reliable predictions of the experimental data. The RELAP5/MOD3.1 overpredicts the heat transfer coefficients of experiment, whereas the RELAP5/MOD3.2 underpredicts those data. It is recommended that the film condensation model in RELAP5/MOD3.2 should be modified to have a larger heat transfer coefficient than those of the present model to give the reliable predictions. 7 refs., 6 figs., 1 tab. (Author)

Assessment of RELAP5/MOD3 with condensation experiment for pure steam condensation in a vercal tube

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Jae; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

The film condensation models in RELAP5/MOD3.1 and RELAP5/MOD3.2 are assessed with the data of experiment performed in the scaled down condensation experimental facility with a single vertical tube of inner diameter of 46 mm in the range of pressure 0.1 {approx} 7.5 MPa for the PSCS(Passive Secondary Condenser System). Both MOD3.1 and MOD3.2 don`t shows any reliable predictions of the experimental data. The RELAP5/MOD3.1 overpredicts the heat transfer coefficients of experiment, whereas the RELAP5/MOD3.2 underpredicts those data. It is recommended that the film condensation model in RELAP5/MOD3.2 should be modified to have a larger heat transfer coefficient than those of the present model to give the reliable predictions. 7 refs., 6 figs., 1 tab. (Author)

Evaluation of fuel-temperature feedback mechanisms in TRAC-PF1/MOD2/NESTLE

International Nuclear Information System (INIS)

Knepper, Paula L.; Feltus, Madeline; Hochreiter, L.E.; Ivanov, Kostadin

1999-01-01

Coupled spatial kinetics and thermal-hydraulics system codes provide a means to model transient nuclear reactor behavior more accurately. Transients marked by strong perturbations, both with thermal-hydraulics and neutronics, such as a control-rod ejection or a main steam-line break, are especially of interest. It is now feasible to model complex reactor behavior with a coupled thermal-hydraulics and spatial kinetics code that provides a means to forecast safety margins. Recently, the Transient Reactor Analysis Code (TRAC)-PF1/MOD2, Version 5.4.25, was coupled with the NESTLE code. This coupled code (TRAC-PF1/MOD2/NESTLE) is used to examine effective fuel-temperature models. The Electric Power Research Institute (EPRI) rod-ejection benchmark was analyzed to evaluate the influence of effective fuel temperature. The rod-ejection transient tests only the fuel-rod, heat-conduction coupling. The coolant thermal-hydraulic coupling is not tested because of the speed of the transient. The neutronics solution changes extremely rapidly, whereas the convective heat transfer at the fuel surface requires more time to influence the coolant temperature of the system. The need to model the response of the system coolant temperature is not crucial in this analysis. The influence of the effective fuel temperature is the key component of this study. Various models were examined using the coupled code to calculate effective fuel temperatures. The influence of different, effective fuel-temperature models on the coupled-code results is studied. Three effective fuel-temperature models are examined: (l) volume average effective fuel temperature, (2) the effective fuel-temperature model suggested by the Office of Economic Cooperation and Development (OECD) rod-ejection benchmark, and (3) the NESTLE effective fuel-temperature model. A discussion is provided describing the effective fuel-temperature models examined in TRAC-PF1/MOD2/NESTLE and the influence of effective fuel temperature in

Upgrade of the RFX-mod real time control system

International Nuclear Information System (INIS)

Manduchi, G.; Barbalace, A.; Luchetta, A.; Soppelsa, A.; Taliercio, C.; Zampiva, E.

2012-01-01

Highlights: ► The paper describes the experience in running the real-time control system of RFX-mod. ► It proposes a new architecture based multicore technology. ► It analyzes two different solutions for data acquisition. ► It discusses the effect of non simultaneous sampling in acquisition. ► It provides some preliminary performance measurements. - Abstract: The real-time control system of RFX-mod, in operation since 2005, has been successful and has allowed several important achievements in the RFX physics research program. As a consequence of this fact, new control algorithms are under investigation, which are more demanding in terms of both enhanced computing power and reduced system latency, currently around 1.5 ms. For this reason, a major upgrade of the system is being considered, and a new architecture has been proposed, taking advantage of the rapid evolution of computer technology in the last years. The central component of the new architecture is a Linux-based multicore server, where individual cores replace the VME computers. The server is connected to the I/O via PCI-e based bus extenders, and every PCI-e connection is managed by a separate core. The system is supervised by MARTe, a software framework for real-time applications written in C++ and developed at JET and currently used for the JET vertical stabilization and in other fusion devices.

Peer review of RELAP5/MOD3 documentation

International Nuclear Information System (INIS)

Craddick, W.G.

1993-01-01

A peer review was performed on a portion of the documentation of the RELAP5/MOD3 computer code. The review was performed in two phases. The first phase was a review of Volume 3, Developmental Assessment problems, and Volume 4, Models and Correlations. The reviewers for this phase were Dr. Peter Griffith, Dr. Yassin Hassan, Dr. Gerald S. Lellouche, Dr. Marino di Marzo and Mr. Mark Wendel. The reviewers recommended a number of improvements, including using a frozen version of the code for assessment guided by a validation plan, better justification for flow regime maps and extension of models beyond their data base. The second phase was a review of Volume 6, Quality Assurance of Numerical Techniques in RELAP5/MOD3. The reviewers for the second phase were Mr. Mark Wendel and Dr. Paul T. Williams. Recommendations included correction of numerous grammatical and typographical errors and better justification for the use of Lax's Equivalence Theorem

RELAP 4/MOD 6 boiling water nodalization study

International Nuclear Information System (INIS)

Sonneck, G.; Pfau, H.

1985-09-01

The risk of nuclear steam supply systems is dominated by the core melt accidents. The first step to a realistic assessment of these sequences is the successful prediction of a loss of coolant event in a test loop. One of the codes for that is RELAP 4/MOD 6 and one of the important options in this code is the nodalization. The base of this work is the test LOCA No. 1 FIX II in Studsvik (Sweden) which also served as the OECD International Standard Problem 15. This report discusses the influence of different nodalizations, of different distributions of pressure, water and structural heat as well as of different bubble rise options, break flow coefficients, and heat transfer time steps. The most important result is that a simple RELAP 4/MOD6 model with less than 10 volumes is able to predict an experiment as LOCA No. 1 in FIX II successfully using only a fraction of the usual computing time. (Author)

RELAP4/MOD-5-CEA pump coastdown experiment simulation

International Nuclear Information System (INIS)

Borges, R.C.; Freitas, R.L.

1988-07-01

Since is important the theoretical-experimental comparison to evaluate the computer codes, these paper presents the simulation with RELAP4/MOD5 Code of a loss of power energy in the pump of the ''Circuito Experimental de Agua-CEA''. From the results attained, the existing models in the Code showed to be very satisfatory quantitative and qualitative behavior of the attained experimental results. (author) [pt

Modification of the bubble rise model used in RELAP4/Mod5 computer code for transients analysis

International Nuclear Information System (INIS)

Scharfmann, E.

1981-01-01

To improve the separation phase and heat transfer models in RELAP4/MOD5 computer code, in order to make more realistic estimates of the thermohydraulic behavior of the core submitted to a loss-of-coolant accident, is the objective of this work. This research is directed to the accident analysis caused by small breaks in the primary circuit of PWR plants, where two-phase flow occurs most of the time. Calculation have been performed with the help of the original version of RELAP code, and the version containing the proposed modifications on this work. Comparing one results with the original ones, we arrive at the conclusion that our results show more conservative values of core pressure and coolant temperature, while the peak values of fuel temperature are not exceeded. (Author) [pt

MOD-5A wind turbine generator program design report: Volume 1: Executive Summary

Science.gov (United States)

1984-01-01

The design, development and analysis of the 7.3 MW MOD-5A wind turbine generator covering work performed between July 1980 and June 1984 is discussed. The report is divided into four volumes: Volume 1 summarizes the entire MOD-5A program, Volume 2 discusses the conceptual and preliminary design phases, Volume 3 describes the final design of the MOD-5A, and Volume 4 contains the drawings and specifications developed for the final design. Volume 1, the Executive Summary, summarizes all phases of the MOD-5A program. The performance and cost of energy generated by the MOD-5A are presented. Each subsystem - the rotor, drivetrain, nacelle, tower and foundation, power generation, and control and instrumentation subsystems - is described briefly. The early phases of the MOD-5A program, during which the design was analyzed and optimized, and new technologies and materials were developed, are discussed. Manufacturing, quality assurance, and safety plans are presented. The volume concludes with an index of volumes 2 and 3.

Evaluation of CNA I coolant channel behaviour during an accidental transient using ICARE2 V2 mod2.3 code

International Nuclear Information System (INIS)

Marino, Edgardo J.L.

1999-01-01

Using the input data language of ICARE2 V2 Mod.3 code, the fuel element and coolant channel assembly of CNA I type was described. This input data was utilized to analyze the system behavior and determine the degradation produced during a hypothetical accidental transient at CNA I. The boundary conditions were determined through a previous calculation with RELAP5/MOD 3.2 code. The results had shown characteristic degradation phenomena's. The temperature of bundle components increases fast after 6.11 h in the first case and 5.28 h in the second case, due to the energy release by cladding oxidation. It was correlated with instantaneous hydrogen production and energy contribution. The cumulated hydrogen production was estimated as 0.15 Kg in the first case and ∼ 5 times greater in the second case. Fission product release from the gap due to cladding rupture took place from 6.25 h in the first case and 5.65 h in the second. Relocation started after 6.81 h in the first case and 5.68 in the second, because the cladding dislocation condition is reached. UO 2 dissolution by molten Zircaloy was observed at different levels in the calculation domain. (author)

RELAP5/MOD3 AP600 problems

International Nuclear Information System (INIS)

Riemke, R.A.

1993-01-01

RELAP5/MOD3 is a reactor systems analysis code that has been developed jointly by the US Nuclear Regulatory Commission (USNRC) and a consortium consisting of several of the countries and domestic organizations that were members of the International Code Assessment and Applications Program (ICAP). The code is currently being used to simulate transients for the next generation of advanced light water reactors (ALWR's). One particular reactor design is the Westinghouse AP600 pressurized water reactor (PWR), which consists of two hot legs and four cold legs as well as passive emergency core cooling (ECC) systems. Initial calculations with RELAP5/MOD3 indicated that the code was not as robust as RELAP5/MOD2.5 with regard to AP600 calculations. Recent modifications in the areas of condensation wall heat transfer, interfacial heat transfer in the presence of noncondensibles, bubbly flow interfacial heat transfer, and time smoothing of both interfacial drag and interfacial heat transfer have improved the robustness, although more reliability is needed

Properties of the periplasmic ModA molybdate-binding protein of Escherichia coli.

Science.gov (United States)

Rech, S; Wolin, C; Gunsalus, R P

1996-02-02

The modABCD operon, located at 17 min on the Escherichia coli chromosome, encodes the protein components of a high affinity molybdate uptake system. Sequence analysis of the modA gene (GenBank L34009) predicts that it encodes a periplasmic binding protein based on the presence of a leader-like sequence at its N terminus. To examine the properties of the ModA protein, the modA structural gene was overexpressed, and its product was purified. The ModA protein was localized to the periplasmic space of the cell, and it was released following a gentle osmotic shock. The N-terminal sequence of ModA confirmed that a leader region of 24 amino acids was removed upon export from the cell. The apparent size of ModA is 31.6 kDa as determined by gel sieve chromatography, whereas it is 22.5 kDa when examined by SDS-polyacrylamide gel electrophoresis. A ligand-dependent protein mobility shift assay was devised using a native polyacrylamide gel electrophoresis protocol to examine binding of molybdate and other anions to the ModA periplasmic protein. Whereas molybdate and tungstate were bound with high affinity (approximately 5 microM), sulfate, chromate, selenate, phosphate, and chlorate did not bind even when tested at 2 mM. A UV spectral assay revealed apparent Kd values of binding for molybdate and tungstate of 3 and 7 microM, respectively. Strains defective in the modA gene were unable to transport molybdate unless high levels of the anion were supplied in the medium. Therefore the modA gene product is essential for high affinity molybdate uptake by the cell. Tungstate interference of molybdate acquisition by the cell is apparently due in part to the high affinity of the ModA protein for this anion.

Transient simulation of ALWR passive safety systems using RELAP5/MOD2

International Nuclear Information System (INIS)

Elias, E.; Nekhamkin, Y.; Arshavski, I.

2004-01-01

Numerical simulation is presented of some passive safety systems currently incorporated in the design of the next generation advanced light water reactors (ALWRs). The performance and effectiveness of ex-core natural convection cooling and the concept of gravity driven water injection at high pressure are investigated using the RELAP5/MOD2 thermal-hydraulic code. The study identifies areas that should be investigated more fully in future experimental programs related to hypothetical large and small LOCA in ALWRs. (author)

Vessel coolant mass depletion during a 5% SBLOCA in the Semiscale Mod-2C facility

International Nuclear Information System (INIS)

Shaw, R.A.; Loomis, G.G.

1985-01-01

Experimental results are presented from two 5% small-break loss-of-coolant accident (SBLOCA) simulations in the Semiscale Mod-2C facility. In performing the simulated 5% SBLOCAs, boundary conditions scaled from a pressurized water reactor (PWR) were used. The experiment was run with initial conditions typical of a PWR (15.6 MPa pressure and 35 K core differential temperature). The Mod-2C facility represents the state-of-the-art in small facilities scaled from PWRs. Phenomena which occurred during the transient included: primary fluid saturation (change from subcooled to saturated blowdown), break uncovery (a centerline break was simulated), condensation-induced liquid hold-up in the steam generator primary tubes, pump suction liquid seal formation and core level depression with resulting core rod temperature excursion, pump suction liquid seal clearance, loop fluid mass redistribution, and gradual core rewet. The influence of core bypass flow is also discussed. 11 refs., 13 figs

Utilization of the RELAP4/MOD5/SAS code version in loss of coolant accident in the Angra 1 nuclear power station

International Nuclear Information System (INIS)

Sabundjian, G.; Freitas, R.L.

1991-09-01

A new version of computer code RELAP4/MOD5 was developed to improve the output. The new version, called RELAP4/MOD5/SAS, prints the main variables in graphical form. In order to check the program, a 36 - volume simulation of the Loss-of-Coolant Accident for Angra - I was performed and the results compared to those of a existing 44 - volume simulation showed a satisfactory agreement with a substantial reduction in computing time. (author)

GeoMod 2014 - Modelling in geoscience

Science.gov (United States)

Leever, Karen; Oncken, Onno

2016-08-01

GeoMod is a biennial conference to review and discuss latest developments in analogue and numerical modelling of lithospheric and mantle deformation. GeoMod2014 took place at the GFZ German Research Centre for Geosciences in Potsdam, Germany. Its focus was on rheology and deformation at a wide range of temporal and spatial scales: from earthquakes to long-term deformation, from micro-structures to orogens and subduction systems. It also addressed volcanotectonics and the interaction between tectonics and surface processes (Elger et al., 2014). The conference was followed by a 2-day short course on "Constitutive Laws: from Observation to Implementation in Models" and a 1-day hands-on tutorial on the ASPECT numerical modelling software.

Verification of the HDR-test V44 using the computer program RALOC-MOD1/83

International Nuclear Information System (INIS)

Jahn, H.; Pham, T. v.; Weber, G.; Pham, B.T.

1985-01-01

RALOC-MOD1/83 was extended by a drainage and sump level modul and several component models to serve as a containment systems code for various LWR types. One such application is to simulate the blowdown in a full pressure containment which is important for the short and long term hydrogen distribution. The post test calculation of the containment standard problem experiment HDR-V44 shows a good agreement, to the test data. The code may be used for short and long term predictions, but it was learned that double containments need the representation of the gap between the inner and outer shell into several zones to achieve a good long-term temperature prediction. The present work completes the development, verification and documentation of RALOC-MOD1. (orig.) [de

TRAC-PF1 MOD1 post test calculations of the OECD LOFT Experiment LP-SB-1

International Nuclear Information System (INIS)

Allen, E.J.

1990-04-01

Analysis of the small, hot leg break, OECD LOFT Experiment LP-SB-1. using the ''best-estimate'' computer code TRAC-PF1/MOD1 is presented. Descriptions of the LOFT facility and the LP-SB-1 experiment are given and development of the TRAC-PF1/MOD1 input model is detailed. The calculations performed in achieving the steady state conditions, from which the experiment was initiated, and the specification of experimental boundary conditions are outlined. 24 refs., 66 figs., 12 tabs

Real-time sensing and gas jet mitigation of VDEs on Alcator C-Mod

Science.gov (United States)

Granetz, R. S.; Wolfe, S. M.; Izzo, V. A.; Reinke, M. L.; Terry, J. L.; Hughes, J. W.; Zhurovich, K.; Whyte, D. G.; Bakhtiari, M.; Wurden, G.

2006-10-01

Experiments have been carried out in Alcator C-Mod to test the effectiveness of gas jet disruption mitigation of VDEs with real-time detection and triggering by the C-Mod digital plasma control system (DPCS). The DPCS continuously computes the error in the plasma vertical position from the magnetics diagnostics. When this error exceeds an adjustable preset value, the DPCS triggers the gas jet valve (with a negligible latency time). The high-pressure gas (argon) only takes a few milliseconds to enter the vacuum chamber and begin affecting the plasma, but this is comparable to the VDE timescale on C-Mod. Nevertheless, gas jet injection reduced the halo current, increased the radiated power fraction, and reduced the heating of the divertor compared to unmitigated disruptions, but not quite as well as in earlier mitigation experiments with vertically stable plasmas. Presumably a faster overall response time would be beneficial, and several ways to achieve this will also be discussed.

Microscopic observation drug susceptibility assay (MODS for early diagnosis of tuberculosis in children.

Directory of Open Access Journals (Sweden)

Dang Thi Minh Ha

2009-12-01

Full Text Available MODS is a novel liquid culture based technique that has been shown to be effective and rapid for early diagnosis of tuberculosis (TB. We evaluated the MODS assay for diagnosis of TB in children in Viet Nam. 217 consecutive samples including sputum (n = 132, gastric fluid (n = 50, CSF (n = 32 and pleural fluid (n = 3 collected from 96 children with suspected TB, were tested by smear, MODS and MGIT. When test results were aggregated by patient, the sensitivity and specificity of smear, MGIT and MODS against "clinical diagnosis" (confirmed and probable groups as the gold standard were 28.2% and 100%, 42.3% and 100%, 39.7% and 94.4%, respectively. The sensitivity of MGIT and MODS was not significantly different in this analysis (P = 0.5, but MGIT was more sensitive than MODS when analysed on the sample level using a marginal model (P = 0.03. The median time to detection of MODS and MGIT were 8 days and 13 days, respectively, and the time to detection was significantly shorter for MODS in samples where both tests were positive (P<0.001. An analysis of time-dependent sensitivity showed that the detection rates were significantly higher for MODS than for MGIT by day 7 or day 14 (P<0.001 and P = 0.04, respectively. MODS is a rapid and sensitive alternative method for the isolation of M.tuberculosis from children.

Preliminary validation of RELAP5/Mod4.0 code for LBE cooled NACIE facility

Energy Technology Data Exchange (ETDEWEB)

Kumari, Indu; Khanna, Ashok, E-mail: akhanna@iitk.ac.in

2017-04-01

Highlights: • Detail discussion of thermo physical properties of Lead Bismuth Eutectic incorporated in the code RELAP5/Mod4.0 included. • Benchmarking of LBE properties in RELAP5/Mod4.0 against literature. • NACIE facility for three different power levels (10.8, 21.7 and 32.5 kW) under natural circulation considered for benchmarking. • Preliminary validation of the LBE properties against experimental data. • NACIE facility for power level 22.5 kW considered for validation. - Abstract: The one-dimensional thermal hydraulic computer code RELAP5 was developed for thermal hydraulic study of light water reactor as well as for nuclear research reactors. The purpose of this work is to evaluate the code RELAP5/Mod4.0 for analysis of research reactors. This paper consists of three major sections. The first section presents detailed discussions on thermo-physical properties of Lead Bismuth Eutectic (LBE) incorporated in RELAP5/Mod4.0 code. In the second section, benchmarking of RELAP5/Mod4.0 has been done with the Natural Circulation Experimental (NACIE) facility in comparison with Barone’s simulations using RELAP5/Mod3.3. Three different power levels (10.8 kW, 21.7 kW and 32.5 kW) under natural circulation conditions are considered. Results obtained for LBE temperatures, temperature difference across heat section, pin surface temperatures, mass flow rates and heat transfer coefficients in heat section heat exchanger are in agreement with Barone’s simulation results within 7% of average relative error. Third section presents validation of RELAP5/Mod4.0 against the experimental data of NACIE facility performed by Tarantino et al. test number 21 at power of 22.5 kW comparing the profiles of temperatures, mass flow rate and velocity of LBE. Simulation and experimental results agree within 7% of average relative error.

ModA and ModB, two ADP-ribosyltransferases encoded by bacteriophage T4: catalytic properties and mutation analysis.

Science.gov (United States)

Tiemann, Bernd; Depping, Reinhard; Gineikiene, Egle; Kaliniene, Laura; Nivinskas, Rimas; Rüger, Wolfgang

2004-11-01

Bacteriophage T4 encodes three ADP-ribosyltransferases, Alt, ModA, and ModB. These enzymes participate in the regulation of the T4 replication cycle by ADP-ribosylating a defined set of host proteins. In order to obtain a better understanding of the phage-host interactions and their consequences for regulating the T4 replication cycle, we studied cloning, overexpression, and characterization of purified ModA and ModB enzymes. Site-directed mutagenesis confirmed that amino acids, as deduced from secondary structure alignments, are indeed decisive for the activity of the enzymes, implying that the transfer reaction follows the Sn1-type reaction scheme proposed for this class of enzymes. In vitro transcription assays performed with Alt- and ModA-modified RNA polymerases demonstrated that the Alt-ribosylated polymerase enhances transcription from T4 early promoters on a T4 DNA template, whereas the transcriptional activity of ModA-modified polymerase, without the participation of T4-encoded auxiliary proteins for middle mode or late transcription, is reduced. The results presented here support the conclusion that ADP-ribosylation of RNA polymerase and of other host proteins allows initial phage-directed mRNA synthesis reactions to escape from host control. In contrast, subsequent modification of the other cellular target proteins limits transcription from phage early genes and participates in redirecting transcription to phage middle and late genes.

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

Upgrade of the RFX-mod real time control system

Energy Technology Data Exchange (ETDEWEB)

Manduchi, G., E-mail: gabriele.manduchi@igi.cnr.it [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, Padova 35127 (Italy); Barbalace, A.; Luchetta, A.; Soppelsa, A.; Taliercio, C.; Zampiva, E. [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, Padova 35127 (Italy)

2012-12-15

Highlights: Black-Right-Pointing-Pointer The paper describes the experience in running the real-time control system of RFX-mod. Black-Right-Pointing-Pointer It proposes a new architecture based multicore technology. Black-Right-Pointing-Pointer It analyzes two different solutions for data acquisition. Black-Right-Pointing-Pointer It discusses the effect of non simultaneous sampling in acquisition. Black-Right-Pointing-Pointer It provides some preliminary performance measurements. - Abstract: The real-time control system of RFX-mod, in operation since 2005, has been successful and has allowed several important achievements in the RFX physics research program. As a consequence of this fact, new control algorithms are under investigation, which are more demanding in terms of both enhanced computing power and reduced system latency, currently around 1.5 ms. For this reason, a major upgrade of the system is being considered, and a new architecture has been proposed, taking advantage of the rapid evolution of computer technology in the last years. The central component of the new architecture is a Linux-based multicore server, where individual cores replace the VME computers. The server is connected to the I/O via PCI-e based bus extenders, and every PCI-e connection is managed by a separate core. The system is supervised by MARTe, a software framework for real-time applications written in C++ and developed at JET and currently used for the JET vertical stabilization and in other fusion devices.

A containment convective loop analysis using the RELAP5-Mod 3.2

International Nuclear Information System (INIS)

Ventura, M.

1996-01-01

The present study was performed to verify the RELAP5-Mod 3.2 code capability to calculate convection phenomena of the type occurring in a convective loop. A simplified geometrical model of a reactor containment system was used. The parametric studies were made for the main variables which govern material transport in the volume junctions considered. The results obtained and that got using the same model with the CONTAIN code, were compared. The comparison is satisfactory. (author). 3 refs., 11 figs

Assessment of RELAP5/MOD3 Version 7 based on the BETHSY Test 6.2 TC

International Nuclear Information System (INIS)

Choi, C.J.; Roth, P.A.; Schultz, R.R.

1992-01-01

This document provides a discussion of the BETHSY test 6.2 TC which was conducted to investigate thermal hydraulic phenomena during a 5% cold leg SBLOCA and to provide high quality data for advanced thermal-hydraulic code assessment. BETHSY test 6.2 TC was analyzed using RELAP5/MOD3 version 7o

RELAP5/MOD2 analysis of LOFT Experiment L9-3

International Nuclear Information System (INIS)

Birchley, J.C.

1992-04-01

An analysis has been performed of LOFT Experiment L9-3, a loss-of-feedwater anticipated transient without trip, in order to support the validation of RELAP5/MOD2. Experiment L9-3 exhibited a rapid boildown of the steam generator, following the loss of feed, with the reactor remaining close to its initial power until the steam generator tubes became sufficiently uncovered for primary to secondary heat transfer to be significantly reduced. The ensuing heat up of the primary fluid resulted in a reduction in power induced by the moderator feedback. The primary system pressure increased to the safety relief valve setpoint, before the fall in reactor power allowed the mismatch between primary system heat input and heat removal via the steam generator to be accommodated by cycling of the pilot operated relief valve (PORV). Comparison between calculation and data shows generally good agreement, though with discrepancies in some areas. Weaknesses in the code's treatment of interphase drag and in the representation of the pressuriser spray are indicated, although a shortage of definitive data, particularly in the steam generator, may also be a factor. The overprediction of interphase drag led to a tendency to underpredict the initial inventory in the steam generator and also, perhaps, to overpredict the steam generator heat transfer while the tubes were being uncovered. There is indication that the pressuriser vapour region conditions were close to equilibrium during spray operation. The point kinetics model in RELAP5/MOD2 proved a viable means of representing the power history for this transient

MELMRK 2.0: A description of computer models and results of code testing

International Nuclear Information System (INIS)

Wittman, R.S.; Denny, V.; Mertol, A.

1992-01-01

An advanced version of the MELMRK computer code has been developed that provides detailed models for conservation of mass, momentum, and thermal energy within relocating streams of molten metallics during meltdown of Savannah River Site (SRS) reactor assemblies. In addition to a mechanistic treatment of transport phenomena within a relocating stream, MELMRK 2.0 retains the MOD1 capability for real-time coupling of the in-depth thermal response of participating assembly heat structure and, further, augments this capability with models for self-heating of relocating melt owing to steam oxidation of metallics and fission product decay power. As was the case for MELMRK 1.0, the MOD2 version offers state-of-the-art numerics for solving coupled sets of nonlinear differential equations. Principal features include application of multi-dimensional Newton-Raphson techniques to accelerate convergence behavior and direct matrix inversion to advance primitive variables from one iterate to the next. Additionally, MELMRK 2.0 provides logical event flags for managing the broad range of code options available for treating such features as (1) coexisting flow regimes, (2) dynamic transitions between flow regimes, and (3) linkages between heatup and relocation code modules. The purpose of this report is to provide a detailed description of the MELMRK 2.0 computer models for melt relocation. Also included are illustrative results for code testing, as well as an integrated calculation for meltdown of a Mark 31a assembly

Compatibility of dip-coated Er2O3 coating by MOD method with liquid Li

International Nuclear Information System (INIS)

Zhang Dongxun; Kondo, Masatoshi; Tanaka, Teruya; Muroga, Takeo; Valentyn, Tsisar

2011-01-01

An electrical insulating ceramic coating on the self-cooled lithium blanket is a promising technology for suppressing MHD pressure drop in the blanket system. Er 2 O 3 is thought to be one of the potential candidate materials for ceramic coatings because of their high electrical resistivity and high compatibility with liquid lithium. In this study, Er 2 O 3 coating was fabricated on the ferritic steels by dip-coating method with MOD (metal organic decomposition) liquid precursor followed by baking in different atmosphere. The coated specimens were immersed at 500 o C in the static liquid lithium to test the compatibility. It was shown that the compatibility of the coating was degraded when Fe 2 O 3 or Fe 3 O 4 was formed as the main composition of the substrate oxidation layer during the baking. On the other hand, thin Cr 2 O 3 layer in the substrate oxidation layer did not influence the stability of Er 2 O 3 coating. Atmosphere controlling for suppressing the substrate oxidation, especially Fe 2 O 3 or Fe 3 O 4 , during the baking is shown to be essential for the compatibility of MOD Er 2 O 3 coating on ferritic steels.

BEACON/MOD: a computer program for thermal-hydraulic analysis of nuclear reactor containments - user's manual

International Nuclear Information System (INIS)

Broadus, C.R.; Doyle, R.J.; James, S.W.; Lime, J.F.; Mings, W.J.

1980-04-01

The BEACON code is a best-estimate, advanced containment code designed to perform a best-estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped-parameter representations for the various parts of the system. The current version of BEACON, which is designated BEACON/MOD3, contains mass and heat transfer models for wall film and wall conduction. It is suitable for the evaluation of short-term transients in dry-containment systems. This manual describes the models employed in BEACON/MOD3 and specifies code implementation requirements. It provides application information for input data preparation and for output data interpretation

GalMod: the last frontier of Galaxy population synthesis models

Science.gov (United States)

Pasetto, Stefano; Kollmeier, Juna; Grebel, Eva K.; chiosi, cesare

2018-01-01

We present a novel Galaxy population synthesis model: GalMod (Pasetto et al. 2016, 2017a,b) is the only star-count model featuring an asymmetric bar/bulge as well as spiral arms as directly obtained by applying linear perturbative theory to self-consistent distribution function of the Galaxy stellar populations. Compared to previous literature models (e.g., Besancon, Trilegal), GalMod allows to generate full-sky mock catalogue, M31 surveys and provides a better match to observed Milky Way (MW) stellar fields.The model can generate synthetic mock catalogs of visible portions of the MW, external galaxies like M31, or N-body simulation initial conditions. At any given time, e.g., a chosen age of the Galaxy, the model contains a sum of discrete stellar populations, namely bulge/bar, disk, halo. The disk population is itself the sum of subpopulations: spiral arms, thin disk, thick disk, and gas component, while the halo is modeled as the sum of a stellar component, a hot coronal gas, and a dark matter component. The Galactic potential is computed from these subpopulations' density profiles and used to generate detailed kinematics by considering the first few moments of the Boltzmann collisionless equation for all the stellar subpopulations. The same density profiles are then used to define the observed color-magnitude diagrams within an input field of view from an arbitrary solar location. Several photometric systems have been included and made available on-line, e.g., SDSS, Gaia, 2MASS, HST WFC3, and others. Finally, we model the extinction with advanced ray tracing solutions.The model's web page (and tutorial) can be accessed at www.GalMod.org.

Framatome's experience in implementing and runnig RELAP5 MOD1

Energy Technology Data Exchange (ETDEWEB)

Truong, T.X.; Rousset, P.

1984-12-01

Implementation of RELAP MOD1 on Framatome's computer began in 1982, when we were working with the electrical power research institute on a safety and relief valve test program. It has been carried out in two stages: a first implementation on the CISI CYBER 740 computer; a transfer of files and a second on implemented our own CYBER 835 computer. The RELAP5 version currently implemented and used in Framatome is the cycle 19 standard version, no modification has been made yet, though some changes in data output files are intended.

Peer review of RELAP5/MOD3 documentation

International Nuclear Information System (INIS)

Craddick, W.G.

1994-01-01

A peer review was performed on a portion of the documentation of the RELAP5/MOD3 computer code. The review was performed in two phases. The first phase was a review of Vol. III, Developmental Assessment Problems, and Vol. IV, Models and Correlations. The reviewers for this phase were Dr. Peter Griffith, Dr. Yassin Hassan, Dr. Gerald S. Lellouche, Dr. Marino di Marzo and Mr. Mark Wendel. The reviewers recommended a number of improvements, including using a frozen version of the code for assessment guided by a validation plan, better discussion of discrepancies between the code and experimental data, and better justification for flow regime maps and extension of models beyond their data base. The second phase was a review of Vol. VI, Quality Assurance of Numerical Techniques in RELAP5/MOD3. The reviewers for the second phase were Mr. Mark Wendel and Dr. Paul T. Williams. Recommendations included correction of numerous grammatical and typographical errors and better justification for the use of Lax's Equivalence Theorem

Characterization of the Hamamatsu 8" R5912-MOD Photomultiplier tube

Science.gov (United States)

Kaptanoglu, Tanner

2018-05-01

Current and future neutrino and direct detection dark matter experiments hope to take advantage of improving technologies in photon detection. Many of these detectors are large, monolithic optical detectors that use relatively low-cost, large-area, and efficient photomultiplier tubes (PMTs). A candidate PMT for future experiments is a newly developed prototype Hamamatsu PMT, the R5912-MOD. In this paper we describe measurements made of the single photoelectron time and charge response of the R5912-MOD, as well as detail some direct comparisons to similar PMTs. Most of these measurements were performed on three R5912-MOD PMTs operating at gains close to 1 × 107. The transit time spread (σ) and the charge peak-to-valley were measured to be on average 680ps and 4.2 respectively. The results of this paper show the R5912-MOD is an excellent candidate for future experiments in several regards, particularly due to its narrow spread in timing.

TRAC-PF1/MOD 1 independent assessment: Semiscale MOD-2A feedwater-line break (S-SF-3) and steam-line break (S-SF-5) tests

International Nuclear Information System (INIS)

Dobranich, D.

1985-11-01

The TRAC-PF1/MOD1 independent assessment project at Sandia is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. As part of this effort, calculations for Semiscale Mod-2A test S-SF-3, a feedwater-line break test, and S-SF-5, a steam-line break test, were performed with TRAC-PF1/MOD1. Most aspects of both the S-SF-3 and S-SF-5 steady-state calculations were found to be in good agreement with data. However, the need for a better steam separator model was identified from the S-SF-3 calculation. Overall, the qualitative behavior of both transients was calculated reasonably well; however, there were some discrepancies in the prediction of the quantitative behavior. The results for the S-SF-3 transient calculation indicate that the primary-to-secondary heat transfer degradation began too early. This was possibly due to overprediction of entrainment in the steam generator boiler, leading to an incorrect calculation of the secondary-side fluid distribution during the steady state. However, there was insufficient data to verify this. Results for the S-SF-5 transient calculation indicate that the primary-side fluid temperature response to a steam-line break was in reasonable agreement with data but the pressure response did not coincide with the data. Uncertainties in the data are sufficient to prevent us from determining the exact cause of this discrepancy, but there is indirect evidence that the calculated rate of phase change in the pressurizer was incorrect. 16 refs., 73 figs., 13 tabs

Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option

International Nuclear Information System (INIS)

Dupleac, D.; Perez, M.; Reventos, F.; Allison, C.

2011-01-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis (IUA) package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). RELAP/SCDAPSIM/MOD4.0(IUA) follows the input-propagation approach using probability distribution functions to define the uncertainty of the input parameters. The main steps for this type of methodologies, often referred as to statistical approaches or Wilks’ methods, are the ones that follow: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. RELAP/SCDAPSIM/MOD4.0(IUA) calculates the number of required code runs given the desired percentile and confidence level, performs the sampling process for the

Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option

Energy Technology Data Exchange (ETDEWEB)

Dupleac, D., E-mail: danieldu@cne.pub.ro [Politehnica Univ. of Bucharest (Romania); Perez, M.; Reventos, F., E-mail: marina.perez@upc.edu, E-mail: francesc.reventos@upc.edu [Technical Univ. of Catalonia (Spain); Allison, C., E-mail: iss@cableone.net [Innovative Systems Software (United States)

2011-07-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis (IUA) package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). RELAP/SCDAPSIM/MOD4.0(IUA) follows the input-propagation approach using probability distribution functions to define the uncertainty of the input parameters. The main steps for this type of methodologies, often referred as to statistical approaches or Wilks’ methods, are the ones that follow: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. RELAP/SCDAPSIM/MOD4.0(IUA) calculates the number of required code runs given the desired percentile and confidence level, performs the sampling process for the

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

Quick Look Report for Semiscale MOD-2C Test S-FS-2

International Nuclear Information System (INIS)

Boucher, T.J.; Chen, T.H.

1985-01-01

Results of a preliminary analysis of the first test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-2 simulated a pressurized water reactor transient initiated by a double-ended offset shear of a steam generator main steam line upstream of the flow restrictor. Initial conditions represented normal ''hot-standby'' operation. The transient included an initial 600-s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant at conditions required to allow a natural circulation cooldown. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overcooling and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overcooling and primary-to-secondary heat transfer. 57 figs., 3 tabs

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

Nonaxisymmetric field effects on Alcator C-Mod

International Nuclear Information System (INIS)

Wolfe, S.M.; Hutchinson, I.H.; Granetz, R.S.; Rice, J.; Hubbard, A.; Lynn, A.; Phillips, P.; Hender, T.C.; Howell, D.F.; La Haye, R.J.; Scoville, J.T.

2005-01-01

A set of external coils (A-coils) capable of producing nonaxisymmetric, predominantly n=1, fields with different toroidal phase and a range of poloidal mode m spectra has been used to determine the threshold amplitude for mode locking over a range of plasma parameters in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda, P. Bonoli, S. Fairfax, C. Fiore, J. Goetz, S. Golovato, R. Granetz, M. Greenwald et al., Phys. Plasmas 1, 1511 (1994)]. The threshold perturbations and parametric scalings, expressed in terms of (B 21 /B T ), are similar to those observed on larger, lower field devices. The threshold is roughly linear in density, with typical magnitudes of order 10 -4 . This result implies that locked modes should not be significantly more problematic for the International Thermonuclear Experimental Reactor [I. P. B. Editors, Nucl. Fusion 39, 2286 (1999)] than for existing devices. Coordinated nondimensional identity experiments on the Joint European Torus [Fusion Technol. 11, 13 (1987)], DIII-D [Fusion Technol. 8, 441 (1985)], and C-Mod, with matching applied mode spectra, have been carried out to determine more definitively the field and size scalings. Locked modes on C-Mod are observed to result in braking of core toroidal rotation, modification of sawtooth activity, and significant reduction in energy and particle confinement, frequently leading to disruptions. Intrinsic error fields inferred from the threshold studies are found to be consistent in amplitude and phase with a comprehensive model of the sources of field errors based on 'as-built' coil and bus-work details and coil imperfections inferred from measurements using in situ magnetic diagnostics on dedicated test pulses. Use of the A-coils to largely cancel the 2/1 component of the intrinsic nonaxisymmetric field has led to expansion of the accessible operating space in C-Mod, including operation up to 2 MA plasma current at 8 T

Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

Energy Technology Data Exchange (ETDEWEB)

D`Auria, F.; Galassi, G.M. [Univ. of Pisa (Italy); Frogheri, M. [Univ. of Genova (Italy)

1997-12-31

The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

Energy Technology Data Exchange (ETDEWEB)

D` Auria, F; Galassi, G M [Univ. of Pisa (Italy); Frogheri, M [Univ. of Genova (Italy)

1998-12-31

The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

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

Post test analysis of TEPSS tests -P2-, -P3-, -P5- and -P7- using the system code RELAP5/MOD 3.2

International Nuclear Information System (INIS)

Luebbesmeyer, D.

2000-01-01

For the PANDA-Test-Facility (TEPSS configuration) post-test calculations and analyses have been performed for experiment -P2- (Early Start), -P3- (PCC start up), -P5- (Symmetric case, Two PCCs only) and -P7- (Severe Accident). Post test calculations have been performed with the system code RELAP5/Mod 3.2 using two different nodalization of the PANDA facility namely a basis nodalization and a much reduced one. The general trend of the calculations can be summarised: RELAP5/Mod3.2 calculated the general trends of the experiments sufficiently accurate; Using the reduced nodalization the results seem to be slightly more accurate than for the basic nodalization; On the other hand, calculations based on the reduced nodalization are not significantly faster than those with basic nodalization; The mass error is in the order of 200 to 900 kg. (author)

CONTEMPT4/MOD6: a multicompartment containment system analysis program

International Nuclear Information System (INIS)

Lin, C.C.; Economos, C.; Lehner, J.R.; Maise, G.

1986-03-01

CONTEMPT4/MOD6 is a digital computer program that describes the response of multicompartment containment system subjected to postulated loss-of-coolant accident (LOCA) conditions. The program is written in FORTRAN IV and can accomodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures and mass and energy inventories due to intercompartment mass and energy exchange taking into account user supplied descriptions of compartments, intercompartment junction flow areas, LOCA source terms and user selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers and BWR pressure suppression systems. To accommodate degraded core type accidents, analytical models for hydrogen and carbon monoxide combustion within compartments and energy transfer due to gas radiation are also provided. Dynamic storage allocation (DSA) is used to limit the amount of computer core used for each problem. The flexibility needed to more realistically model the complexity of prototypical containments is provided by the multicompartment capability (up to 999 individual compartments) and generalized user oriented input data descriptions. The program employs an implicit algorithm to compute junction flow when numerically induced flow oscillations are encountered. This capability provides significant reduction of computer run time relative to previous codes in the CONTEMPT series. Descriptions of these analytical models are presented, together with input instructions for the CONTEMPT4/MOD6 program and sample problem results. 23 refs., 62 figs

A plan for the modification and assessment of TRAC-PF1/MOD2 for use in analyzing CANDU 3 transient thermal-hydraulic phenomena

International Nuclear Information System (INIS)

Siebe, D.A.; Boyack, B.E.; Giguere, P.T.

1994-11-01

This report presents the results of the review and planning done for the United States Nuclear Regulatory Commission to identify the thermal-hydraulic phenomena that could occur in the CANDU 3 reactor design during transient conditions, plan modifications to the TRAC-PF1/MOD2 (TRAC) computer code needed to adequately predict CANDU 3 transient thermal-hydraulic phenomena, and identify an assessment program to verify the ability of TRAC, when modified, to predict these phenomena. This work builds on analyses and recommendations produced by the Idaho National Engineering Laboratory (INEL). To identify the thermal-hydraulic phenomena, a large-break loss-of-coolant accident simulation, performed as part of earlier work by INEL with an Atomic Energy of Canada, Limited (AECL) thermal-hydraulic computer code (CATHENA), was analyzed in detail. Other accident scenarios were examined for additional phenomena. A group of Los Alamos National Laboratory reactor thermal-hydraulics experts ranked the phenomena to produce a preliminary phenomena identification and ranking table (PIRT). The preliminary nature of the PIRT was a result of a lack of direct expertise with the unique processes and phenomena of the CANDU 3. Nonetheless, this PIRT provided an adequate foundation for planning a program of code modifications. We believe that this PIRT captured the most important phenomena and that refinements to the PIRT will mainly produce clarification of the relative importance (ranking) of phenomena. A plan for code modifications was developed based on this PIRT and on information about the modeling methodologies for CANDU-specific phenomena used in AECL codes. AECL thermal-hydraulic test facilities and programs were reviewed and the information used in developing an assessment plan to ensure that TRAC-PF1/MOD2, when modified, will adequately predict CANDU 3 phenomena

Post-test analysis of PIPER-ONE PO-IC-2 experiment by RELAP5/MOD3 codes

International Nuclear Information System (INIS)

Bovalini, R.; D'Auria, F.; Galassi, G.M.; Mazzini, M.

1996-11-01

RELAP5/MOD3.1 was applied to the PO-IC-2 experiment performed in PIPER-ONE facility, which has been modified to reproduce typical isolation condenser thermal-hydraulic conditions. RELAP5 is a well known code widely used at the University of Pisa during the past seven years. RELAP5/MOD3.1 was the latest version of the code made available by the Idaho National Engineering Laboratory at the time of the reported study. PIPER-ONE is an experimental facility simulating a General Electric BWR-6 with volume and height scaling ratios of 1/2,200 and 1./1, respectively. In the frame of the present activity a once-through heat exchanger immersed in a pool of ambient temperature water, installed approximately 10 m above the core, was utilized to reproduce qualitatively the phenomenologies expected for the Isolation Condenser in the simplified BWR (SBWR). The PO-IC-2 experiment is the flood up of the PO-SD-8 and has been designed to solve some of the problems encountered in the analysis of the PO-SD-8 experiment. A very wide analysis is presented hereafter including the use of different code versions

Self-initialization module for the RELAP4/MOD7 program

International Nuclear Information System (INIS)

Behling, S.R.; Burgess, C.H.; Johnsen, G.W.

1979-01-01

An automated self-initialization option has been developed and implemented into the RELAP4/MOD7 computer code. With this new feature, pressurized water reactors or similar models may be initialized to steady state conditions with minimal user-supplied input data. Previously, the analyst (user) had to hand compute all pressures and temperatures in conjunction with several iterative computer runs. Two semi-independent functions, pressure balancing and thermal balancing, comprise the self-initialization module. The user supplies input data specifying a reference pressure and temperature and associated locations. Control volume pressures and temperatures are computed consistent with the nodalization, geometry, total power, and mass flow rates specified. Verification has been accomplished by performing transient calculations on system models and observing the persistence of stable operating conditions. Preparation of input data for the self-initialization module seldom requires more than several man-hours. Computing time required on a CDC 7600 computer to complete the self-initialization process is typically less than 30 seconds

Alcator C-MOD proposal addendum

International Nuclear Information System (INIS)

Bonoli, P.; Greenwald, M.; Gwinn, D.

1986-04-01

Since the design concept and overall purpose of the Alcator C-MOD device are similar to that proposed in October 1985, we have chosen in this document only to highlight areas where changes or additions have been made. Chapters in the Addendum correspond to those in the Proposal, except Chapter 9 which describes a number of toroidal improvement concepts which are being considered for inclusion in the Alcator C-MOD experimental program. A description of the redesign and a discussion of the objectives of the experimental program are given

Bascule d'un modèle poutre à un modèle 3D en dynamique des machines tournantes

OpenAIRE

Tannous , Mikhael; Cartraud , Patrice; Dureisseix , David; Torkhani , Mohamed

2013-01-01

National audience; Les problèmes de machines tournantes incluant un contact rotor-stator, nécessitent un maillage 3D de la zone de contact. Cependant, un modèle 3D pour toute la durée de simulation conduit à des temps de calcul rédhibitoires. Or un modèle poutre est suffisant pour décrire la dynamique de la machine tournante hors contact. Une stratégie qui permet d'utiliser un modèle poutre et un autre 3D, pendant deux phases différentes durant la même simulation, permet donc de gagner en tem...

Preparation and characterization of Al{sub 2}O{sub 3} coating by MOD method on CLF-1 RAFM steel

Energy Technology Data Exchange (ETDEWEB)

Wang, L. [Key Laboratory of Radiation Physics and Technology (Sichuan University), Ministry of Education, Institute of Nuclear Science and Technology, Sichuan University, Chengdu 610064 (China); Yang, J.J., E-mail: jjyang@scu.edu.cn [Key Laboratory of Radiation Physics and Technology (Sichuan University), Ministry of Education, Institute of Nuclear Science and Technology, Sichuan University, Chengdu 610064 (China); Feng, Y.J. [Southwestern Institute of Physics, Chengdu 614000 (China); Li, F.Z.; Liao, J.L.; Yang, Y.Y. [Key Laboratory of Radiation Physics and Technology (Sichuan University), Ministry of Education, Institute of Nuclear Science and Technology, Sichuan University, Chengdu 610064 (China); Feng, K.M. [Southwestern Institute of Physics, Chengdu 614000 (China); Liu, N. [Key Laboratory of Radiation Physics and Technology (Sichuan University), Ministry of Education, Institute of Nuclear Science and Technology, Sichuan University, Chengdu 610064 (China)

2017-04-15

Metal organic decomposition (MOD) method was proposed to prepare Al{sub 2}O{sub 3} TPB coatings on CLF-1 RAFM steel. A comprehensive characterization of SEM, XPS, and XRD demonstrated the formation of Al{sub 2}O{sub 3} coatings. The effect of the preparation parameters, including annealing temperature T{sub a}, withdrawal speed V{sub w} and immersion time t{sub i} on the microstructure and properties of the coatings was investigated. It showed that amorphous aluminum oxide coating began to transform to γ-Al{sub 2}O{sub 3} at temperature of T{sub a} = 600 °C. The Al{sub 2}O{sub 3} coating with T{sub a} = 700 °C and T{sub b} = 500 °C performed the best crystallization feature. The hardness of the coatings gradually increased with increasing V{sub w}, while the corrosion resistance exhibited a reverse trend. Meanwhile, the nanohardness and corrosion resistance of the coating with t{sub i} = 300 s was improved as compared to the coating with t{sub i} = 0 s. Moreover, the effect of particle size and substrate oxidation on the mechanical property and corrosion resistance of the coatings was discussed. - Highlights: •MOD method was proposed to prepare Al{sub 2}O{sub 3} TPB on CLF-1 RAFM steel. •Effect of preparation parameters on the coating microstructure and properties was studied preliminary. •High quality MOD coating can be developed by multi-baking process.

Radon transport modelling: User's guide to RnMod3d

International Nuclear Information System (INIS)

Andersen, C.E.

2000-08-01

RnMod3d is a numerical computer model of soil-gas and radon transport in porous media. It can be used, for example, to study radon entry from soil into houses in response to indoor-outdoor pressure differences or changes in atmospheric pressure. It can also be used for flux calculations of radon from the soil surface or to model radon exhalation from building materials such as concrete. The finite-volume model is a technical research tool, and it cannot be used meaningfully without good understanding of the involved physical equations. Some understanding of numerical mathematics and the programming language Pascal is also required. Originally, the code was developed for internal use at Risoe only. With this guide, however, it should be possible for others to use the model. Three-dimensional steady-state or transient problems with Darcy flow of soil gas and combined generation, radioactive decay, diffusion and advection of radon can be solved. Moisture is included in the model, and partitioning of radon between air, water and soil grains (adsorption) is taken into account. Most parameters can change in time and space, and transport parameters (diffusivity and permeability) may be anisotropic. This guide includes benchmark tests based on simple problems with known solutions. RnMod3d has also been part of an international model intercomparison exercise based on more complicated problems without known solutions. All tests show that RnMod3d gives results of good quality. (au)

An Optimization Scheme for ProdMod

International Nuclear Information System (INIS)

Gregory, M.V.

1999-01-01

A general purpose dynamic optimization scheme has been devised in conjunction with the ProdMod simulator. The optimization scheme is suitable for the Savannah River Site (SRS) High Level Waste (HLW) complex operations, and able to handle different types of optimizations such as linear, nonlinear, etc. The optimization is performed in the stand-alone FORTRAN based optimization deliver, while the optimizer is interfaced with the ProdMod simulator for flow of information between the two

Transport and confinement studies in the RFX-mod reversed-field pinch experiment

International Nuclear Information System (INIS)

Innocente, P.; Alfier, A.; Carraro, L.; Lorenzini, R.; Pasqualotto, R.; Terranova, D.

2007-01-01

In the modified RFX experiment (RFX-mod) external magnetic field coils and a close fitting thin conductive shell control radial magnetic fields. In the so-called virtual shell (VS) operation, radial field zeroing at the thin shell radius is stationary provided by the feedback-controlled coils. First experiments on RFX-mod proved the capability of the active scheme to steadily reduce the radial magnetic field. Furthermore it has been found that such edge magnetic field control extends its beneficial effects to the whole plasma. With respect to the old RFX, where magnetohydrodynamic modes amplitude was controlled by the use of a passive thick conductive shell, a stationary 2- to 3-fold reduction of the B r field amplitude in the core is obtained. The reduction of field fluctuations positively reflects on confinement. In fact, a strong reduction of the loop voltage is observed and correspondingly a 3-fold increase in pulse length is achieved by using the same poloidal flux swing. Temperature and particle measurements confirm the improved confinement properties of the VS operation. With a lower ohmic input power, higher electron temperature and lower particle influx are measured. Particle and heat transport have been studied by means of a 1D code. Local power balance was used to compute the heat conductivity profile: for the VS discharges a lower conductivity over a significant region of the plasma is found. The improved properties of RFX-mod VS operation provide a better confinement scaling in terms of plasma current. The results show that compared with the thick shell configuration, a significant confinement improvement can be obtained under stationary conditions by actively controlling the plasma magnetic boundary

A new formulation of the law of octic reciprocity for primes ≡±3(mod8 and its consequences

Directory of Open Access Journals (Sweden)

Richard H. Hudson

1982-01-01

Full Text Available Let p and q be odd primes with q≡±3(mod8, p≡1(mod8=a2+b2=c2+d2 and with the signs of a and c chosen so that a≡c≡1(mod4. In this paper we show step-by-step how to easily obtain for large q necessary and sufficient criteria to have (−1(q−1/2q(p−1/8≡(a−bd/acj(modp for j=1,…,8 (the cases with j odd have been treated only recently [3] in connection with the sign ambiguity in Jacobsthal sums of order 4. This is accomplished by breaking the formula of A.E. Western into three distinct parts involving two polynomials and a Legendre symbol; the latter condition restricts the validity of the method presented in section 2 to primes q≡3(mod8 and significant modification is needed to obtain similar results for q≡±1(mod8. Only recently the author has completely resolved the case q≡5(mod8, j=1,…,8 and a sketch of the method appears in the closing section of this paper.

Fabrication of High-Quality SmBa2Cu3O7-δ Thin Films by a Modified TFA-MOD Process

International Nuclear Information System (INIS)

Kim, Duck Jin; Moon, Seung Hyun; Park, Chan; Yoo, Sang Im; Song, Kyu Jeong

2005-01-01

We report a successful fabrication of high-quality SmBa 2 Cu 3 O 7-δ (SmBCO) thin films on LaAlO 3 (LAO)(100) single crystalline substrates by a modified TFA-MOD method. After the pyrolysis heat treatment of spin-coated films up to 400 degree C, SmBCO films were fired at various temperatures ranging from 810 to 850 degree C in a reduced oxygen atmosphere (10 ppm O 2 in Ar). Optimally processed SmBCO films exhibited the zero-resistance temperature (T c ,zero) of 90.2 K and the critical current density (J c ) of 0.8 MA/cm 2 at 77K in self-field. Compared with the J c values (normally, > 2 MA/cm 2 at 77 K) of MOD-TFA processed YBCO films, rather depressed J c values in SmBCO films are most probably attributed to the existence of alpha-axis oriented grains.

Watershed Modeling Applications with the Open-Access Modular Distributed Watershed Educational Toolbox (MOD-WET) and Introductory Hydrology Textbook

Science.gov (United States)

Huning, L. S.; Margulis, S. A.

2014-12-01

Traditionally, introductory hydrology courses focus on hydrologic processes as independent or semi-independent concepts that are ultimately integrated into a watershed model near the end of the term. When an "off-the-shelf" watershed model is introduced in the curriculum, this approach can result in a potential disconnect between process-based hydrology and the inherent interconnectivity of processes within the water cycle. In order to curb this and reduce the learning curve associated with applying hydrologic concepts to complex real-world problems, we developed the open-access Modular Distributed Watershed Educational Toolbox (MOD-WET). The user-friendly, MATLAB-based toolbox contains the same physical equations for hydrological processes (i.e. precipitation, snow, radiation, evaporation, unsaturated flow, infiltration, groundwater, and runoff) that are presented in the companion e-textbook (http://aqua.seas.ucla.edu/margulis_intro_to_hydro_textbook.html) and taught in the classroom. The modular toolbox functions can be used by students to study individual hydrologic processes. These functions are integrated together to form a simple spatially-distributed watershed model, which reinforces a holistic understanding of how hydrologic processes are interconnected and modeled. Therefore when watershed modeling is introduced, students are already familiar with the fundamental building blocks that have been unified in the MOD-WET model. Extensive effort has been placed on the development of a highly modular and well-documented code that can be run on a personal computer within the commonly-used MATLAB environment. MOD-WET was designed to: 1) increase the qualitative and quantitative understanding of hydrological processes at the basin-scale and demonstrate how they vary with watershed properties, 2) emphasize applications of hydrologic concepts rather than computer programming, 3) elucidate the underlying physical processes that can often be obscured with a complicated

Experiment data report for Semiscale Mod-1 Test S-05-2 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Collins, B.L.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-2 of the Semiscale Mod-1 alternate emergency core coolant (ECC) injection test series. This test is one of 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-05-2 was conducted from an initial cold leg fluid temperature of 545 0 F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 65 0 F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux of nuclear fuel rods during a loss-of-coolant accident

MOD-RTG multicouple test results and mission readiness

International Nuclear Information System (INIS)

Hartman, R.F.; Kelly, C.E.

1993-01-01

MOD-RTG represents the design configuration for the next generation of Radioisotope Thermoelectric Generators (RTG), aimed at improving specific power and efficiency over current General Purpose Heat Source Radioisotope Thermoelectric Generators (GPHS-RTGs). The modular RTG reference design has been described in previous papers (Hartman 1988). The multicouple is a key element required for the successful development of the modular RTG. The multicouple is a high voltage, thermoelectric device employing a close packed, glass bonded thermopile array of twenty thermoelectric couples, connected in a series circuit. The multicouple is designed to operate at a 1270 K hot junction temperature and a 570 K cold junction temperature, yielding a power output of approximately 2.1 watts at 3.5 volts at beginning of life. The objectives of the MOD-RTG program are focused on establishing a multicouple life test data base and life prediction capability which will permit, with reasonable margin, a projected multicouple life of greater than eight (8) years. This paper summarizes the current status of multicouple life testing and performance modeling and the level of technology readiness needed to demonstrate mission readiness for MOD-RTG

A CO2 laser polarimeter for measurement of plasma current profile in Alcator C-Mod

International Nuclear Information System (INIS)

Ma, C.H.; Hutchinson, D.P.; Richards, R.K.; Irby, J.; Luke, T.

1994-01-01

A multichannel infrared polarimeter system for measurement of the plasma current profile in Alcator C-Mod has been designed, constructed, and tested. The system utilizes a cw CO 2 , laser at a wavelength of 10.6 μm. An electro-optic polarization-modulation technique has been used to achieve the high sensitivity required for the measurement. The recent results of the measurements as well as the feasibility of its application on ITER are presented

Gas jet disruption mitigation studies on Alcator C-Mod

International Nuclear Information System (INIS)

Granetz, R.; Whyte, D.G.; Izzo, V.A.; Biewer, T.; Reinke, M.L.; Terry, J.; Bader, A.; Bakhtiari, M.; Jernigan, T.; Wurden, G.

2006-01-01

Damaging effects of disruptions are a major concern for Alcator C-Mod, ITER and future tokamak reactors. High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the operational requirements of fast response time and reliability, while still being benign to subsequent discharges. Disruption mitigation experiments using an optimized gas jet injection system are being carried out on Alcator C-Mod to study the physics of gas jet penetration into high pressure plasmas, as well as the ability of the gas jet impurities to convert plasma energy into radiation on timescales consistent with C-Mod's fast quench times, and to reduce halo currents given C-Mod's high-current density. The dependence of impurity penetration and effectiveness on noble gas species (He, Ne, Ar, Kr) is also being studied. It is found that the high-pressure neutral gas jet does not penetrate deeply into the C-Mod plasma, and yet prompt core thermal quenches are observed on all gas jet shots. 3D MHD modelling of the disruption physics with NIMROD shows that edge cooling of the plasma triggers fast growing tearing modes which rapidly produce a stochastic region in the core of the plasma and loss of thermal energy. This may explain the apparent effectiveness of the gas jet in C-Mod despite its limited penetration. The higher-Z gases (Ne, Ar, Kr) also proved effective at reducing halo currents and decreasing thermal deposition to the divertor surfaces. In addition, noble gas jet injection proved to be benign for plasma operation with C-Mod's metal (Mo) wall, actually improving the reliability of the startup in the following discharge

Assessment of RELAP5/MOD2 code using loss of offsite power transient data of KNU [Korea Nuclear Unit] No. 1 Plant

International Nuclear Information System (INIS)

Chung, Bud-Dong; Kim, Hho-Jung

1990-04-01

This report presents a code assessment study based on a real plant transient that occurred on June 9, 1981 at the KNU number-sign 1 (Korea Nuclear Unit Number 1). KNU number-sign 1 is a two-loop Westinghouse PWR plant of 587 Mwe. The loss of offsite power transient occurred at the 77.5% reactor power with 0.5%/hr power ramp. The real plant data were collected from available on-line plant records and computer diagnostics. The transient was simulated by RELAP5/MOD2/36.05 and the results were compared with the plant data to assess the code weaknesses and strengths. Some nodalization studies were performed to contribute to developing a guideline for PWR nodalization for the transient analysis. 5 refs., 18 figs., 3 tabs

Alcator C-Mod predictive modeling

International Nuclear Information System (INIS)

Pankin, Alexei; Bateman, Glenn; Kritz, Arnold; Greenwald, Martin; Snipes, Joseph; Fredian, Thomas

2001-01-01

Predictive simulations for the Alcator C-mod tokamak [I. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] are carried out using the BALDUR integrated modeling code [C. E. Singer et al., Comput. Phys. Commun. 49, 275 (1988)]. The results are obtained for temperature and density profiles using the Multi-Mode transport model [G. Bateman et al., Phys. Plasmas 5, 1793 (1998)] as well as the mixed-Bohm/gyro-Bohm transport model [M. Erba et al., Plasma Phys. Controlled Fusion 39, 261 (1997)]. The simulated discharges are characterized by very high plasma density in both low and high modes of confinement. The predicted profiles for each of the transport models match the experimental data about equally well in spite of the fact that the two models have different dimensionless scalings. Average relative rms deviations are less than 8% for the electron density profiles and 16% for the electron and ion temperature profiles

Heat Transfer treatment in computer codes for safety analysis

International Nuclear Information System (INIS)

Jerele, A.; Gregoric, M.

1984-01-01

Increased number of operating nuclear power plants has stressed importance of nuclear safety evaluation. For this reason, accordingly to regulatory commission request, safety analyses with computer codes are preformed. In this paper part of this thermohydraulic models dealing with wall-to-fluid heat transfer correlations in computer codes TRAC=PF1, RELAP4/MOD5, RELAP5/MOD1 and COBRA-IV is discussed. (author)

Analysis of VVER-1000 large and small break LOCA experiments with RELAP5/MOD3.2

International Nuclear Information System (INIS)

Rychkov, M.; Chikkanagoudar, U.; Sehgal, B.R.

2004-01-01

A RELAP5 model for the analysis of the PSB-VVER test facility was developed by EREC in Russia. The PSB-VVER is a large-scale integral test facility to model the VVER-1000 type NPP. The volume and power scale in this test facility is 1:300 and the elevation scale is 1:1, which corresponds to the elevation mark of the reactor prototype. At the Division of Nuclear Safety, we have modified the PSB-VVER facility's RELAP5 model in order to analyze two of the transient tests performed on the PSB-VVER facility, which serve as the validation matrix described by NEA/CSNI. The objective of the work conducted was to validate the results obtained from RELAP5's calculation with the supplied experimental data from the PSB-VVER test facility. Two accident scenarios have been calculated and analyzed. After being verified against the '11% UP LOCA' test data, the RELAP5/MOD3.2 model was used for a so-called 'blind' transient calculation of the test '2*25% HL LOCA' and the results obtained were compared with the experimental data provided after the calculation. From the results of the qualitative and quantitative comparison of the 2 test calculations and the experimental data, we can state that the RELAP5/MOD3.2 code gives a satisfactory modeling of the PSB-VVER facility' thermal hydraulic phenomena

Modifications of the bubble rise model and heat transfer model used in RELAP 4/Mod 5 computer code for transient analysis

International Nuclear Information System (INIS)

Scharfmann, E.; Silva, D.E. da

1981-01-01

The modifications on the phase separation model and heat tranfer model in Relap4/Mod 5 computer code, in order to make more realistic estimates of the core thermohydraulic behavior submitted to a loss of coolant accident. This research is directed to the accident analysis caused by small breaks in the primary circuits of PWR plants, where two-phase flow occurs most of the time. Calculation have been performed with the help of the original version of Relap code, as well as the version containing the proposed modifications on this work. Comparing one results with the original ones, we arrive at the conclusion that our results show more conservative values of core pressure and coolant temperature, while the peak values of fuel temperature are not exceeded. (Author) [pt

Modélisation par éléments finis 3D du champ magnétostatique dans les enroulements des réactances cuirassées de grande puissance. Comparaison avec le calcul en 2D

Science.gov (United States)

Ngnegueu, Triomphant; Terme, Claude; Mailhot, Michel

1993-03-01

In this paper, the finite element method is applied for the computation of the magnetostatic field in the windings of a shell-form reactor. The modeling is carried out in 3D, using FLUX3D, a software developed at the Laboratoire d'Electrotechnique de Grenoble. The results are compared to those obtained in 2D. These calculation results are also compared to some test results. Dans cet article, nous décrivons une application de la méthode des éléments finis pour la modélisation du champ magnétostatique dans les enroulements d'une réactance cuirassée de grande puissance. La modélisation est conduite en 3D, en utilisant le logiciel FLUX3D. Les résultats du calcul sont comparés avec ceux obtenus en 2D. Quelques comparaisons sont aussi effectuées avec des résultats de mesure.

Comparisons of TRAC-PD2 calculations with Semiscale Mod-3 small-break tests

International Nuclear Information System (INIS)

Gilbert, J.S.; Sahota, M.S.; Boyack, B.E.; Booker, C.P.; Meier, J.K.

1981-01-01

Five experiments conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory (INEL) were calculated using the latest released version of the Transient Reactor Analysis Code (TRAC-PD2). The results were used to assess TRAC-PD2 predictions of thermal-hydraulic phenomena and the effects of pump operation on system response during slow transients. Tests S-SB-P1, S-SB-P2, and S-SB-P7 simulated equivalent 2.5% communicative cold-leg breaks for early pump-trip (pumps-off), intermediate pump-trip (pumps-on), and late pump-trip (pumps-on) operation, respectively. Tests S-SB-P3 and S-SB-P4 simulated equivalent 2.5% communicative hot-leg breaks for pumps-off and pumps-on operation, respectively. Parameters examined in the study included primary system mass distribution, mass inventory, and void fraction distribution

Twenty Years of Research on the Alcator C-Mod Tokamak

Science.gov (United States)

Greenwald, Martin

2013-10-01

Alcator C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since its start in 1993, contributing data that extended tests of critical physical models into new parameter ranges and into new regimes. Using only RF for heating and current drive with innovative launching structures, C-Mod operates routinely at very high power densities. Research highlights include direct experimental observation of ICRF mode-conversion, ICRF flow drive, demonstration of Lower-Hybrid current drive at ITER-like densities and fields and, using a set of powerful new diagnostics, extensive validation of advanced RF codes. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components--an approach adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and EDA H-mode regimes which have high performance without large ELMs and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and found that self-generated flow shear can be strong enough to significantly modify transport. C-Mod made the first quantitative link between pedestal temperature and H-mode performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. Work supported by U.S. DoE

Assessment of RELAP5/MOD3.1 using LOFT L2-3 experiment data

International Nuclear Information System (INIS)

Lee, Sang Yong; Ban, Chang Hwan; Chung, Bob Dong

1994-06-01

The capability of RELAP5/MOD3.1 to predict overall LOCA thermal hydraulic phenomena was assessed utilizing the data of LOFT L2-3 experiment. Loop behaviors such as mass flow rate, water density, momentum flux, and the heating-up and rewetting of the fuel rod cladding during blowdown were well calculated. Reflood heat-up of the fuel rod cladding at the high power region of the core was reasonably predicted. But in the upper part of the core, cladding heat-up was calculated incorrectly since present code has no capability to calculate the top-down quenching which of highly multi-dimensional behavior. (Author) 10 refs., 46 figs., 2 tabs

Neutral particle diagnostics for ALCATOR C-Mod

International Nuclear Information System (INIS)

Kurz, C.; Fiore, C.L.

1990-01-01

The ALCATOR C-Mod experiment will be equipped with two PPPL charge exchange neutral particle analyzers (CENAs), one of which views the plasma tangentially (R tan /R 0 =1.05), whereas the second has a horizontally scannable sight line (0≤R tan /R 0 ≤0.51). The perpendicularly viewing CENA will be capable of analyzing neutrals up to 600 keV amu for up to three separate species simultaneously. Thus high-energy tails can be observed together with the bulk ion temperature. The operation of both analyzers will allow simultaneous measurements from both the perpendicular and tangential chords. The CENAs will be used to study the effect of ICRF heating on the ion energy distribution with emphasis on the high-energy tail. A Fokker--Planck code (FPPRF) [Hammett, Ph.D. thesis, Princeton (1986)] is used to assess the appropriate operating regime of the analyzer (n≤4x10 20 m -3 for T i =2 keV, for Maxwellian ion energy distribution). The experimental design and computer simulations will be detailed

Experimental/theoretical comparisons of the turbulence in the scrape-off-layers of Alcator C-Mod, DIII-D, and NSTX

International Nuclear Information System (INIS)

Terry, J.L. . E-mail : terry@psfc.mit.edu; Zweben, S.J.; Rudakov, D.L.

2003-01-01

The intermittent turbulent transport in the scrape-off-layers of Alcator C-Mod, DIII-D, and NSTX is studied experimentally. On DIII-D the fluctuations of both density and temperature have strongly non-Gaussian statistics, and events with amplitudes above 10 times the mean level are responsible for large fractions of the net particle and heat transport, indicating the importance of turbulence on the transport. In C-Mod and NSTX the turbulence is imaged with a very high density of spatial measurements. The 2-D structure and dynamics of emission from a localized gas puff are observed, and intermittent features (also sometimes called 'blobs') are typically seen. On DIII-D the turbulence is imaged using BES and similar intermittent features are seen. The dynamics of these intermittent features are discussed. The experimental observations are compared with numerical simulations of edge turbulence. The electromagnetic turbulence in a 3-D geometry is computed using non-linear plasma fluid equations. The wavenumber spectra in the poloidal dimension of the simulations are in reasonable agreement with those of the C-Mod experimental images once the response of the optical system is accounted for. The resistive ballooning mode is the dominant linear instability in the simulations. (author)

Mod-5A wind turbine generator program design report. Volume 4: Drawings and specifications, book 2

Science.gov (United States)

1984-01-01

The design, development and analysis of the 7.3 MW MOD-5A wind turbine generator is documented. There are four volumes. This volume contains the drawings and specifications that were developed in preparation for building the MOD-5A wind turbine generator. This is the second book of volume four. Some of the items it contains are specs for the emergency shutdown panel, specs for the simulator software, simulator hardware specs, site operator terminal requirements, control data system requirements, software project management plan, elastomeric teeter bearing requirement specs, specs for the controls electronic cabinet, and specs for bolt pretensioning.

Evaluation of the MODS culture technique for the diagnosis of tuberculous meningitis.

Directory of Open Access Journals (Sweden)

Maxine Caws

2007-11-01

Full Text Available Tuberculous meningitis (TBM is a devastating condition. The rapid instigation of appropraite chemotherapy is vital to reduce morbidity and mortality. However rapid diagnosis remains elusive; smear microscopy has extremely low sensitivity on cerebrospinal fluid (CSF in most laboratories and PCR requires expertise with advanced infrastructure and has sensitivity of only around 60% under optimal conditions. Neither technique allows for the microbiological isolation of M. tuberculosis and subsequent drug susceptibility testing. We evaluated the recently developed microscopic observation drug susceptibility (MODS assay format for speed and accuracy in diagnosing TBM.Two hundred and thirty consecutive CSF samples collected from 156 patients clinically suspected of TBM on presentation at a tertiary referal hospital in Vietnam were enrolled into the study over a five month period and tested by Ziehl-Neelsen (ZN smear, MODS, Mycobacterial growth Indicator tube (MGIT and Lowenstein-Jensen (LJ culture. Sixty-one samples were from patients already on TB therapy for >1day and 19 samples were excluded due to untraceable patient records. One hundred and fifty samples from 137 newly presenting patients remained. Forty-two percent (n = 57/137 of patients were deemed to have TBM by clinical diagnostic and microbiological criteria (excluding MODS. Sensitivity by patient against clinical gold standard for ZN smear, MODS MGIT and LJ were 52.6%, 64.9%, 70.2% and 70.2%, respectively. Specificity of all microbiological techniques was 100%. Positive and negative predictive values for MODS were 100% and 78.7%, respectively for HIV infected patients and 100% and 82.1% for HIV negative patients. The median time to positive was 6 days (interquartile range 5-7, significantly faster than MGIT at 15.5 days (interquartile range 12-24, and LJ at 24 days (interquartile range 18-35 days (P<0.01.We have shown MODS to be a sensitive, rapid technique for the diagnosis of TBM with

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

International Nuclear Information System (INIS)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

2015-01-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

2015-07-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3

International Nuclear Information System (INIS)

Preece, R.J.; Putney, J.M.

1993-07-01

A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser

Physician-initiated courtesy MODS testing for TB and MDR-TB diagnosis and patient management.

Science.gov (United States)

Nic Fhogartaigh, C J; Vargas-Prada, S; Huancaré, V; Lopez, S; Rodríguez, J; Moore, D A J

2008-05-01

Laboratorio de Investigación de Enfermedades Infecciosas, Universidad Peruana Cayetano Heredia (UPCH) and government health centres, Lima, Peru. To evaluate the contribution of unselected (courtesy) microscopic observation drug susceptibility (MODS) testing to the diagnosis and/or drug susceptibility testing (DST) of tuberculosis and their subsequent impact upon patient management. Retrospective database analysis and case note review of MODS culture-positive cases. Mycobacterium tuberculosis was isolated in 28.9% of 225 samples (209 patients); 22.2% of 63 positive cases were multidrug-resistant. In 58 MODS culture-positive cases with follow-up data available, MODS provided culture confirmation of diagnosis, DST or both in 82.8%, before any standard method. In 41.4%, this result should have prompted a modification in patient management. Delays between laboratory result and initiation or change of treatment, where applicable, took on average 42 and 64 days, respectively, of which a delay of respectively 17 and 48 days occurred after the receipt of results by the health facility. MODS provides important data for clinical management within a meaningful timeframe and should contribute positively to patient outcomes due to earlier initiation of appropriate therapy. Although clinicians may successfully select patients likely to benefit from MODS, ongoing work is required to identify optimal implementation of the assay and to reduce logistical and health system derived delays.

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

Assessment of critical flow models of RELAP5-MOD2 and CATHARE codes

International Nuclear Information System (INIS)

Hao Laomi; Zhu Zhanchuan

1992-01-01

The critical flow tests for the long and short nozzles conducted on the SUPER MOBY-DICK facility were analyzed using the RELAP5-MOD2 and CATHARE 1.3 codes to assess the critical flow models of two codes. The critical mass flux calculated for two nozzles are given. The CATHARE code has used the thermodynamic nonequilibrium sound velocity of the two-phase fluid as the critical flow criterion, and has the better interphase transfer models and calculates the critical flow velocities with the completely implicit solution. Therefore, it can well calculate the critical flowrate and can describe the effect of the geometry L/D on the critical flowrate

20 years of research on the Alcator C-Mod tokamaka)

Science.gov (United States)

Greenwald, M.; Bader, A.; Baek, S.; Bakhtiari, M.; Barnard, H.; Beck, W.; Bergerson, W.; Bespamyatnov, I.; Bonoli, P.; Brower, D.; Brunner, D.; Burke, W.; Candy, J.; Churchill, M.; Cziegler, I.; Diallo, A.; Dominguez, A.; Duval, B.; Edlund, E.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Garcia, O.; Gao, C.; Goetz, J.; Golfinopoulos, T.; Granetz, R.; Grulke, O.; Hartwig, Z.; Horne, S.; Howard, N.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J.; Izzo, V.; Kessel, C.; LaBombard, B.; Lau, C.; Li, C.; Lin, Y.; Lipschultz, B.; Loarte, A.; Marmar, E.; Mazurenko, A.; McCracken, G.; McDermott, R.; Meneghini, O.; Mikkelsen, D.; Mossessian, D.; Mumgaard, R.; Myra, J.; Nelson-Melby, E.; Ochoukov, R.; Olynyk, G.; Parker, R.; Pitcher, S.; Podpaly, Y.; Porkolab, M.; Reinke, M.; Rice, J.; Rowan, W.; Schmidt, A.; Scott, S.; Shiraiwa, S.; Sierchio, J.; Smick, N.; Snipes, J. A.; Snyder, P.; Sorbom, B.; Stillerman, J.; Sung, C.; Takase, Y.; Tang, V.; Terry, J.; Terry, D.; Theiler, C.; Tronchin-James, A.; Tsujii, N.; Vieira, R.; Walk, J.; Wallace, G.; White, A.; Whyte, D.; Wilson, J.; Wolfe, S.; Wright, G.; Wright, J.; Wukitch, S.; Zweben, S.

2014-11-01

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental

Teaching Mods with Class

DEFF Research Database (Denmark)

Champion, Erik

2012-01-01

from around the world, representing fields as diverse as architecture, ethnography, puppetry, cultural studies, music education, interaction design and industrial design. How can we design, play with and reflect on the contribution of game mods, related tools and techniques, to both game studies...

Explorations in computing an introduction to computer science

CERN Document Server

Conery, John S

2010-01-01

Introduction Computation The Limits of Computation Algorithms A Laboratory for Computational ExperimentsThe Ruby WorkbenchIntroducing Ruby and the RubyLabs environment for computational experimentsInteractive Ruby Numbers Variables Methods RubyLabs The Sieve of EratosthenesAn algorithm for finding prime numbersThe Sieve Algorithm The mod Operator Containers Iterators Boolean Values and the delete if Method Exploring the Algorithm The sieve Method A Better Sieve Experiments with the Sieve A Journey of a Thousand MilesIteration as a strategy for solving computational problemsSearching and Sortin

Interactive graphical analyzer based on RELAP5/MOD3.2-NPA

International Nuclear Information System (INIS)

Posada, J.M.; Martin, M.; Reventos, F.; Llopis, C.

1999-01-01

The work presented in this paper consists on the development of a Graphical Interactive Analyzer for Asco (two units) and Vandellos (one unit) Nuclear Power Plants, all of them are three loop Westinghouse PWR with rated electrical power around 1000 Mwe. Basic steps are: Development of the thermal-hydraulic and kinetic model for RELAP5/mod3.2 corresponding to NSSS, Steam Flow paths from Steam Generators to Turbine and Condenser, Feedwater System, Emergency Core Cooling System; and related protection and control systems. Development of Graphical representation, for NPA-1.3.4., to permit the user interact with the model. Validation against experimental data. The result is an engineering tool that can help on Plant transient analysis, and on the study of modifications proposed on the components simulated; it's also a powerful tool for operator teaching. (author)

Second Generation HTs Wire Based on RABiTS Substrates and MOD YBCO

Energy Technology Data Exchange (ETDEWEB)

Schoop, U. [American Superconductor Corporation, Westborough, MA; Rupich, Marty [American Superconductor Corporation, Westborough, MA; Thieme, C. L. H. [American Superconductor Corporation, Westborough, MA; Verebelyi, D. T. [American Superconductor Corporation, Westborough, MA; Zhang, W. [American Superconductor Corporation, Westborough, MA; Li, Xiaoping [American Superconductor Corporation, Westborough, MA; Kodenkandath, Thomas [American Superconductor Corporation, Westborough, MA; Nguyen, N. [American Superconductor Corporation, Westborough, MA; Siegal, E. E. [American Superconductor Corporation, Westborough, MA; Civale, L. [Los Alamos National Laboratory (LANL); Holesinger, T. G. [Los Alamos National Laboratory (LANL); Maiorov, B. [Los Alamos National Laboratory (LANL); Goyal, Amit [ORNL; Paranthaman, Mariappan Parans [ORNL

2005-01-01

The performance of Second Generation (2G) high temperature superconducting wire manufactured by continuous reel-to-reel processes is nearing the 300 A/cm-width (77 K, self field) performance threshold for commercial power cable applications. The 2G manufacturing approach under development at American Superconductor is based on the combination of the RABiTS substrate-buffer technology with metal organic deposition (MOD) of the YBCO layer. The capability of this process has been demonstrated in multiple 10 meter lengths with critical currents exceeding 250 A/cm-width with high uniformity and reproducibility. Critical currents of 380 A/cm-width have been achieved in short length samples prepared by the same basic process. The incorporation of nanoparticles ('nanodots') into the YBCO layer using the MOD process has resulted in a 2-fold improvement in the critical current at 65 K in a 3 T field. The research and development focus at ASMC is now directed toward the economical scale-up of the RABiTS/MOD process, optimization of the conductor properties for targeted applications and the use of 2G wire in initial demonstration applications.

The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

Energy Technology Data Exchange (ETDEWEB)

Agafonova, N. [St. Petersburg State Technical Univ. (Russian Federation); Banati, J. [Lappeenranta Univ. of Technology (Finland)

1997-12-31

RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

Energy Technology Data Exchange (ETDEWEB)

Agafonova, N [St. Petersburg State Technical Univ. (Russian Federation); Banati, J [Lappeenranta Univ. of Technology (Finland)

1998-12-31

RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

Experiment data report for semiscale Mod-1 test S-04-2 (baseline ECC test)

International Nuclear Information System (INIS)

Crapo, H.S.; Collins, B.L.; Sackett, K.E.

1976-09-01

Recorded test data are presented for Test S-04-2 of the Semiscale Mod-1 Baseline ECC test series. This test is among Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system. Test S-04-2 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using emergency core coolant injection parameters based on downcomer volume scaling. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such sime that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core coolant injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown. The purpose of the report is to make available the uninterpreted data from Test S-04-2 for future data analysis and test results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent

Web based electronic logbook and experiment run database viewer for Alcator C-Mod

International Nuclear Information System (INIS)

Fredian, T.W.; Stillerman, J.A.

2006-01-01

Since 1991, the scientists and engineers at the Alcator C-Mod experiment at MIT have been recording text entries about the experiments being performed in an electronic logbook. In addition, separate documents such as run plans, run summaries and experimental proposals have been created and stored in a variety of formats in computer files. This information has now been organized and made available via any modern web browser. The new web based interface permits the user to browse through all the logbook entries, run information and even view some key data traces of the experiment. Since this information is being catalogued by Internet search engines, these tools can also be used to quickly locate information. The web based logbook and run information interface provides some additional capabilities. Once logged into the web site, users can add, delete or modify logbook entries directly from their browser. The logbook window on their browser also provides dynamic updating when any new logbook entries are made. There is also live C-Mod operation status information with optional audio announcements available. The user can receive the same state change announcements such as 'entering init' or 'entering pulse' as they would if they were sitting in the C-Mod control room. This paper will describe the functionality of the web based logbook and how it was implemented

Radon transport modelling: User's guide to RnMod3d

Energy Technology Data Exchange (ETDEWEB)

Andersen, C.E

2000-08-01

RnMod3d is a numerical computer model of soil-gas and radon transport in porous media. It can be used, for example, to study radon entry from soil into houses in response to indoor-outdoor pressure differences or changes in atmospheric pressure. It can also be used for flux calculations of radon from the soil surface or to model radon exhalation from building materials such as concrete. The finite-volume model is a technical research tool, and it cannot be used meaningfully without good understanding of the involved physical equations. Some understanding of numerical mathematics and the programming language Pascal is also required. Originally, the code was developed for internal use at Risoe only. With this guide, however, it should be possible for others to use the model. Three-dimensional steady-state or transient problems with Darcy flow of soil gas and combined generation, radioactive decay, diffusion and advection of radon can be solved. Moisture is included in the model, and partitioning of radon between air, water and soil grains (adsorption) is taken into account. Most parameters can change in time and space, and transport parameters (diffusivity and permeability) may be anisotropic. This guide includes benchmark tests based on simple problems with known solutions. RnMod3d has also been part of an international model intercomparison exercise based on more complicated problems without known solutions. All tests show that RnMod3d gives results of good quality. (au)

Substance use associated with short sleep duration in patients with schizophrenia or schizoaffective disorder.

Science.gov (United States)

Tang, Vivian K; Pato, Michele T; Sobell, Janet L; Hammond, Terese C; Valdez, Mark M; Lane, Christianne J; Pato, Carlos N

2016-06-01

To examine the association between substance use and short sleep duration in individuals with schizophrenia or schizoaffective disorder, depressive type (SADD). Cross-sectional, retrospective study. Urban, suburban, and rural centers across the United States. 2,462 consented, adult individuals with schizophrenia or schizoaffective disorder, depressive type (SADD). Participants included inpatients in acute or chronic care settings as well as outpatients and residents in community dwellings. Substance use was assessed with 10 questions adopted from well-validated measures (e.g., CAGE questionnaire) for alcohol, marijuana, and illicit drugs. Short sleep duration was defined as <6 hr of self-reported sleep per night. Close to 100% of our sample used nicotine while 83% used substances other than nicotine. More importantly, there was a significant association between substance use and short sleep duration. Interestingly, this association was strongest among African-Americans with schizophrenia or SADD. Because psychiatric medications often target chemical receptors involved with both sleep and substance use, understanding the association between short sleep duration and substance use in individuals with schizophrenia and SADD is important. Given that the majority of premature deaths in individuals with psychotic illness are due to medical conditions associated with modifiable risk factors, prospective studies designed to examine the effect of short sleep duration on behaviors like substance use should be undertaken. Finally, analyzing genetic and environmental data in a future study might help illuminate the strong association found between short sleep duration and substance use in African-Americans with schizophrenia and SADD. © 2015 Wiley Periodicals, Inc. © 2015 Wiley Periodicals, Inc.

Københavns Kommunes indsats mod social dumping - målopfyldelsesevaluering

DEFF Research Database (Denmark)

Baadsgaard, Kelvin; Jørgensen, Henning

2016-01-01

Evaluering af, om de politiske intentioner med indsats mod social dumping i Københavns Kommune er blevet indfriet......Evaluering af, om de politiske intentioner med indsats mod social dumping i Københavns Kommune er blevet indfriet...

Evaluating the Auto-MODS Assay, a Novel Tool for Tuberculosis Diagnosis for Use in Resource-Limited Settings

Science.gov (United States)

Wang, Linwei; Mohammad, Sohaib H.; Li, Qiaozhi; Rienthong, Somsak; Rienthong, Dhanida; Nedsuwan, Supalert; Mahasirimongkol, Surakameth; Yasui, Yutaka

2014-01-01

There is an urgent need for simple, rapid, and affordable diagnostic tests for tuberculosis (TB) to combat the great burden of the disease in developing countries. The microscopic observation drug susceptibility assay (MODS) is a promising tool to fill this need, but it is not widely used due to concerns regarding its biosafety and efficiency. This study evaluated the automated MODS (Auto-MODS), which operates on principles similar to those of MODS but with several key modifications, making it an appealing alternative to MODS in resource-limited settings. In the operational setting of Chiang Rai, Thailand, we compared the performance of Auto-MODS with the gold standard liquid culture method in Thailand, mycobacterial growth indicator tube (MGIT) 960 plus the SD Bioline TB Ag MPT64 test, in terms of accuracy and efficiency in differentiating TB and non-TB samples as well as distinguishing TB and multidrug-resistant (MDR) TB samples. Sputum samples from clinically diagnosed TB and non-TB subjects across 17 hospitals in Chiang Rai were consecutively collected from May 2011 to September 2012. A total of 360 samples were available for evaluation, of which 221 (61.4%) were positive and 139 (38.6%) were negative for mycobacterial cultures according to MGIT 960. Of the 221 true-positive samples, Auto-MODS identified 212 as positive and 9 as negative (sensitivity, 95.9%; 95% confidence interval [CI], 92.4% to 98.1%). Of the 139 true-negative samples, Auto-MODS identified 135 as negative and 4 as positive (specificity, 97.1%; 95% CI, 92.8% to 99.2%). The median time to culture positivity was 10 days, with an interquartile range of 8 to 13 days for Auto-MODS. Auto-MODS is an effective and cost-sensitive alternative diagnostic tool for TB diagnosis in resource-limited settings. PMID:25378569

From the Last Interglacial to the Anthropocene: Modelling a Complete Glacial Cycle (PalMod)

Science.gov (United States)

Brücher, Tim; Latif, Mojib

2017-04-01

We will give a short overview and update on the current status of the national climate modelling initiative PalMod (Paleo Modelling, www.palmod.de). PalMod focuses on the understanding of the climate system dynamics and its variability during the last glacial cycle. The initiative is funded by the German Federal Ministry of Education and Research (BMBF) and its specific topics are: (i) to identify and quantify the relative contributions of the fundamental processes which determined the Earth's climate trajectory and variability during the last glacial cycle, (ii) to simulate with comprehensive Earth System Models (ESMs) the climate from the peak of the last interglacial - the Eemian warm period - up to the present, including the changes in the spectrum of variability, and (iii) to assess possible future climate trajectories beyond this century during the next millennia with sophisticated ESMs tested in such a way. The research is intended to be conducted over a period of 10 years, but with shorter funding cycles. PalMod kicked off in February 2016. The first phase focuses on the last deglaciation (app. the last 23.000 years). From the ESM perspective PalMod pushes forward model development by coupling ESM with dynamical ice sheet models. Computer scientists work on speeding up climate models using different concepts (like parallelisation in time) and one working group is dedicated to perform a comprehensive data synthesis to validate model performance. The envisioned approach is innovative in three respects. First, the consortium aims at simulating a full glacial cycle in transient mode and with comprehensive ESMs which allow full interactions between the physical and biogeochemical components of the Earth system, including ice sheets. Second, we shall address climate variability during the last glacial cycle on a large range of time scales, from interannual to multi-millennial, and attempt to quantify the relative contributions of external forcing and processes

Disruption Warning Database Development and Exploratory Machine Learning Studies on Alcator C-Mod

Science.gov (United States)

Montes, Kevin; Rea, Cristina; Granetz, Robert

2017-10-01

A database of about 1800 shots from the 2015 campaign on the Alcator C-Mod tokamak is assembled, including disruptive and non-disruptive discharges. The database consists of 40 relevant plasma parameters with data taken from 160k time slices. In order to investigate the possibility of developing a robust disruption prediction algorithm that is tokamak-independent, we focused machine learning studies on a subset of dimensionless parameters such as βp, n /nG , etc. The Random Forests machine learning algorithm provides insight on the available data set by ranking the relative importance of the input features. Its application on the C-Mod database, however, reveals that virtually no one parameter has more importance than any other, and that its classification algorithm has a low rate of successfully predicted samples, as well as poor false positive and false negative rates. Comparing the analysis of this algorithm on the C-Mod database with its application to a similar database on DIII-D, we conclude that disruption prediction may not be feasible on C-Mod. This conclusion is supported by empirical observations that most C-Mod disruptions are caused by radiative collapse due to molybdenum from the first wall, which happens on just a 1-2ms timescale. Supported by the US Dept. of Energy under DE-FC02-99ER54512 and DE-FC02-04ER54698.

20 years of research on the Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Greenwald, M.; Baek, S.; Barnard, H.; Beck, W.; Bonoli, P.; Brunner, D.; Burke, W.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Gao, C.; Golfinopoulos, T.; Granetz, R.; Hartwig, Z.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J.

2014-01-01

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental

Assessment of RELAP5/MOD2 against critical flow data from Marviken tests JIT 11 and CFT 21

International Nuclear Information System (INIS)

Rosdahl, O.; Caraher, D.

1986-09-01

RELAP5/MOD2 simulations of the critical flow of saturated steam are reported together with simulations of the critical flow of subcooled liquid and a low quality two-phase mixture. The experiments which were simulated used nozzle diameters of 0.3 m and 0.5 m. RELAP5 overpredicted the experimental flow rates by 10 to 25% unless discharge coefficients were applied

Feasibility of using auto Mod-MPI system, a novel technique for automated measurement of fetal modified myocardial performance index.

Science.gov (United States)

Lee, M-Y; Won, H-S; Jeon, E-J; Yoon, H C; Choi, J Y; Hong, S J; Kim, M-J

2014-06-01

To evaluate the reproducibility of measurement of the fetal left modified myocardial performance index (Mod-MPI) determined using a novel automated system. This was a prospective study of 116 ultrasound examinations from 110 normal singleton pregnancies at 12 + 1 to 37 + 1 weeks' gestation. Two experienced operators each measured the left Mod-MPI twice manually and twice automatically using the Auto Mod-MPI system. Intra- and interoperator reproducibility were assessed using intraclass correlation coefficients (ICCs) and the manual and automated measurements obtained by the more experienced operator were compared using Bland-Altman plots and ICCs. Both operators successfully measured the left Mod-MPI in all cases using the Auto Mod-MPI system. For both operators, intraoperator reproducibility was higher when performing automated measurements (ICC = 0.967 and 0.962 for Operators 1 and 2, respectively) than when performing manual measurements (ICC = 0.857 and 0.856 for Operators 1 and 2, respectively). Interoperator agreement was also better for automated than for manual measurements (ICC = 0.930 vs 0.723, respectively). There was good agreement between the automated and manual values measured by the more experienced operator. The Auto Mod-MPI system is a reliable technique for measuring fetal left Mod-MPI and demonstrates excellent reproducibility. Copyright © 2013 ISUOG. Published by John Wiley & Sons Ltd.

Vaccine mod halthed testes i besætning

DEFF Research Database (Denmark)

Lauritsen, Klara Tølbøll

2012-01-01

Ny vaccine mod ledbetændelse forårsaget af Mycoplasma hyosynoviae testes nu hos 200 svin i en problembesætning. Håbet er færre halte svin og en nedbringelse af antibiotikaforbruget.......Ny vaccine mod ledbetændelse forårsaget af Mycoplasma hyosynoviae testes nu hos 200 svin i en problembesætning. Håbet er færre halte svin og en nedbringelse af antibiotikaforbruget....

Functional overview of the Production Planning Model (ProdMod)

International Nuclear Information System (INIS)

Gregory, M.V.; Paul, P.K.

1995-09-01

The Production Planning Model (ProdMod) has been developed by SRTC for use by High Level Waste Program Management and High Level Waste Engineering as a fast running, integrated, comprehensive model of the entire SRS high level waste (HLW) complex. ProdMod can simulate the response of the HLW complex from its current state to the end of tank clean-up or to any intermediate point. The present document describes the initial release of ProdMod at the end of FY95: a model version that contains all the significant elements from the High-level Waste System Plan Revision 5 and is capable of running the simulation all the way to the postulated completion of waste removal. For the scenario represented by this release, that simulates approximately 70 years of operation of the HLW complex (out to FY2065). This initial release of ProdMod will serve as the immediate starting point for the modeling of the High-Level Waste System Plan Revision 6. Thus ProdMod is expected to be in a state of continuous change and improvement.the initial goal has been to generate a simulation of the processes of interest, with the emphasis on mass and volume balances tracked throughout the HLW complex. That has been accomplished. Future development will add a set of cost equations to the process equations and extend the model for use as a linear programming (optimization) application. The goal of this later phase will be to free the ProdMod user to some extent from the need to set up detailed simulation scenarios: the model will automatically make operational choices which minimize or maximize a given objective function. Appendix A contains the source code

Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

Science.gov (United States)

Brunner, D.; Burke, W.; Kuang, A. Q.; LaBombard, B.; Lipschultz, B.; Wolfe, S.

2016-02-01

Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

Modding a free and open source software video game: "Play testing is hard work"

Directory of Open Access Journals (Sweden)

Giacomo Poderi

2014-03-01

Full Text Available Video game modding is a form of fan productivity in contemporary participatory culture. We see modding as an important way in which modders experience and conceptualize their work. By focusing on modding in a free and open source software video game, we analyze the practice of modding and the way it changes modders' relationship with their object of interest. The modders' involvement is not always associated with fun and creativity. Indeed, activities such as play testing often undermine these dimensions of modding. We present a case study of modding that is based on ethnographic research done for The Battle for Wesnoth, a free and open source software strategy video game entirely developed by a community of volunteers.

Ny vaccine mod ledbetændelse er ikke effektiv

DEFF Research Database (Denmark)

Nielsen, Elisabeth Okholm; Lauritsen, Klara Tølbøll

2013-01-01

En ny mulighed for at vaccinere mod mykoplasma-ledbetændelse er undersøgt hos en slagtesvineproducent. Vaccinen kunne desværre ikke forebygge halthed eff ektivt.......En ny mulighed for at vaccinere mod mykoplasma-ledbetændelse er undersøgt hos en slagtesvineproducent. Vaccinen kunne desværre ikke forebygge halthed eff ektivt....

Reliability of the MODS assay decentralisation process in three health regions in Peru

Science.gov (United States)

Mendoza, A.; Castillo, E.; Gamarra, N.; Huamán, T.; Perea, M.; Monroi, Y.; Salazar, R.; Coronel, J.; Acurio, M.; Obregón, G.; Roper, M.; Bonilla, C.; Asencios, L.; Moore, D. A. J.

2011-01-01

OBJECTIVE To deliver rapid isoniazid (INH) and rifampicin (RMP) drug susceptibility testing (DST) close to the patient, we designed a decentralisation process for the microscopic observation drug susceptibility (MODS) assay in Peru and evaluated its reliability. METHODS After 2 weeks of training, laboratory staff processed ≥120 consecutive sputum samples each in three regional laboratories. Samples were processed in parallel with MODS testing at an expert laboratory. Blinded paired results were independently analysed by the Instituto Nacional de Salud (INS) according to predetermined criteria: concordance for culture, DST against INH and RMP and diagnosis of multidrug-resistant t uberculosis (MDR-TB) ≥ 95%, McNemar's P > 0.05, kappa index (κ) ≥ 0.75 and contamination 1–4%. Sensitivity and specificity for MDR-TB were calculated. RESULTS The accreditation process for Callao (126 samples, 79.4% smear-positive), Lima Sur (n = 130, 84%) and Arequipa (n = 126, 80%) took respectively 94, 97 and 173 days. Pre-determined criteria in all regional laboratories were above expected values. The sensitivity and specificity for detecting MDR-TB in regional laboratories were >95%, except for sensitivity in Lima Sur, which was 91.7%. Contamination was 1.0–2.3%. Mean delay to positive MODS results was 9.9–12.9 days. CONCLUSION Technology transfer of MODS was reliable, effective and fast, enabling the INS to accredit regional laboratories swiftly. PMID:21219684

MOD2SEA: A Coupled Atmosphere-Hydro-Optical Model for the Retrieval of Chlorophyll-a from Remote Sensing Observations in Complex Turbid Waters

Directory of Open Access Journals (Sweden)

Behnaz Arabi

2016-09-01

Full Text Available An accurate estimation of the chlorophyll-a (Chla concentration is crucial for water quality monitoring and is highly desired by various government agencies and environmental groups. However, using satellite observations for Chla estimation remains problematic over coastal waters due to their optical complexity and the critical atmospheric correction. In this study, we coupled an atmospheric and a water optical model for the simultaneous atmospheric correction and retrieval of Chla in the complex waters of the Wadden Sea. This coupled model called MOD2SEA combines simulations from the MODerate resolution atmospheric TRANsmission model (MODTRAN and the two-stream radiative transfer hydro-optical model 2SeaColor. The accuracy of the coupled MOD2SEA model was validated using a matchup data set of MERIS (MEdium Resolution Imaging SpectRometer observations and four years of concurrent ground truth measurements (2007–2010 at the NIOZ jetty location in the Dutch part of the Wadden Sea. The results showed that MERIS-derived Chla from MOD2SEA explained the variations of measured Chla with a determination coefficient of R2 = 0.88 and a RMSE of 3.32 mg·m−3, which means a significant improvement in comparison with the standard MERIS Case 2 regional (C2R processor. The proposed coupled model might be used to generate a time series of reliable Chla maps, which is of profound importance for the assessment of causes and consequences of long-term phenological changes of Chla in the turbid Wadden Sea area.

Irradiation hardening of Mod.9Cr-1Mo steel

International Nuclear Information System (INIS)

Ryu, Woo-Seog; Kim, Sung-Ho; Choo, Kee-Nam; Kim, Do-Sik

2009-01-01

An irradiation test of Mod.9Cr-1Mo steel was carried out in the OR5 test hole of HANARO of a 30 MW thermal power at 390±10degC up to a fast neutron fluence of 4.4x10 19 (n/cm 2 ) (E > 1.0 MeV). The dpa of the irradiated specimens was evaluated to be 0.034 - 0.07. Tensile and impact tests of the irradiated Mod.9Cr-1Mo were done in the hot cell of the IMEF. The change of the tensile strength by irradiation was similar to the change of the yield strength. The increase of the yield and tensile strengths was up to 18% and 10% respectively. The elongation reduction of the weldment was up to 65%. (author)

idRHa+ProMod - Rail Hardening Control System

International Nuclear Information System (INIS)

Ferro, L

2016-01-01

idRHa+ProMod is the process control system developed by Primetals Technologies to foresee the thermo-mechanical evolution and micro-structural composition of rail steels subjected to slack quenching into idRHa+ Rail Hardening equipments in a simulation environment. This tool can be used both off-line or in-line, giving the user the chance to test and study the best cooling strategies or letting the automatic control system free to adjust the proper cooling recipe. Optimization criteria have been tailored in order to determine the best cooling conditions according to the metallurgical requirements imposed by the main rail standards and also taking into account the elastoplastic bending phenomena occurring during all stages of the head hardening process. The computational core of idRHa+ProMod is a thermal finite element procedure coupled with special algorithms developed to work out the main thermo-physical properties of steel, to predict the non-isothermal austenite decomposition into all the relevant phases and subsequently to evaluate the amount of latent heat of transformation released, the compound thermal expansion coefficient and the amount of plastic deformation in the material. Air mist and air blades boundary conditions have been carefully investigated by means of pilot plant tests aimed to study the jet impingement on rail surfaces and the cooling efficiency at all working conditions. Heat transfer coefficients have been further checked and adjusted directly on field during commissioning. idRHa+ is a trademark of Primetals Technologies Italy Srl (paper)

idRHa+ProMod - Rail Hardening Control System

Science.gov (United States)

Ferro, L.

2016-03-01

idRHa+ProMod is the process control system developed by Primetals Technologies to foresee the thermo-mechanical evolution and micro-structural composition of rail steels subjected to slack quenching into idRHa+ Rail Hardening equipments in a simulation environment. This tool can be used both off-line or in-line, giving the user the chance to test and study the best cooling strategies or letting the automatic control system free to adjust the proper cooling recipe. Optimization criteria have been tailored in order to determine the best cooling conditions according to the metallurgical requirements imposed by the main rail standards and also taking into account the elastoplastic bending phenomena occurring during all stages of the head hardening process. The computational core of idRHa+ProMod is a thermal finite element procedure coupled with special algorithms developed to work out the main thermo-physical properties of steel, to predict the non-isothermal austenite decomposition into all the relevant phases and subsequently to evaluate the amount of latent heat of transformation released, the compound thermal expansion coefficient and the amount of plastic deformation in the material. Air mist and air blades boundary conditions have been carefully investigated by means of pilot plant tests aimed to study the jet impingement on rail surfaces and the cooling efficiency at all working conditions. Heat transfer coefficients have been further checked and adjusted directly on field during commissioning. idRHa+ is a trademark of Primetals Technologies Italy Srl

An assessment of the annular flow transition criteria and interphase friction models in RELAP5/MOD2

International Nuclear Information System (INIS)

Putney, J.M.

1989-02-01

An assessment of the annular flow transition criteria and interphase friction models for two-phase flow in tubes used in RELAP5/MOD2 code is described. The assessment examines the theoretical bases for the criteria and models and considers the results of comparisons with experimental data. Several deficiencies in the transition criteria are identified and appropriate improvements proposed. The interphase friction models are found to be adequate for PWR analyses. (author)

Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

International Nuclear Information System (INIS)

Tuunanen, J.

1993-01-01

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

Energy Technology Data Exchange (ETDEWEB)

Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

1981-01-01

FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

Identification of functional elements and regulatory circuits by Drosophila modENCODE

Energy Technology Data Exchange (ETDEWEB)

Roy, Sushmita; Ernst, Jason; Kharchenko, Peter V.; Kheradpour, Pouya; Negre, Nicolas; Eaton, Matthew L.; Landolin, Jane M.; Bristow, Christopher A.; Ma, Lijia; Lin, Michael F.; Washietl, Stefan; Arshinoff, Bradley I.; Ay, Ferhat; Meyer, Patrick E.; Robine, Nicolas; Washington, Nicole L.; Stefano, Luisa Di; Berezikov, Eugene; Brown, Christopher D.; Candeias, Rogerio; Carlson, Joseph W.; Carr, Adrian; Jungreis, Irwin; Marbach, Daniel; Sealfon, Rachel; Tolstorukov, Michael Y.; Will, Sebastian; Alekseyenko, Artyom A.; Artieri, Carlo; Booth, Benjamin W.; Brooks, Angela N.; Dai, Qi; Davis, Carrie A.; Duff, Michael O.; Feng, Xin; Gorchakov, Andrey A.; Gu, Tingting; Henikoff, Jorja G.; Kapranov, Philipp; Li, Renhua; MacAlpine, Heather K.; Malone, John; Minoda, Aki; Nordman, Jared; Okamura, Katsutomo; Perry, Marc; Powell, Sara K.; Riddle, Nicole C.; Sakai, Akiko; Samsonova, Anastasia; Sandler, Jeremy E.; Schwartz, Yuri B.; Sher, Noa; Spokony, Rebecca; Sturgill, David; van Baren, Marijke; Wan, Kenneth H.; Yang, Li; Yu, Charles; Feingold, Elise; Good, Peter; Guyer, Mark; Lowdon, Rebecca; Ahmad, Kami; Andrews, Justen; Berger, Bonnie; Brenner, Steven E.; Brent, Michael R.; Cherbas, Lucy; Elgin, Sarah C. R.; Gingeras, Thomas R.; Grossman, Robert; Hoskins, Roger A.; Kaufman, Thomas C.; Kent, William; Kuroda, Mitzi I.; Orr-Weaver, Terry; Perrimon, Norbert; Pirrotta, Vincenzo; Posakony, James W.; Ren, Bing; Russell, Steven; Cherbas, Peter; Graveley, Brenton R.; Lewis, Suzanna; Micklem, Gos; Oliver, Brian; Park, Peter J.; Celniker, Susan E.; Henikoff, Steven; Karpen, Gary H.; Lai, Eric C.; MacAlpine, David M.; Stein, Lincoln D.; White, Kevin P.; Kellis, Manolis

2010-12-22

To gain insight into how genomic information is translated into cellular and developmental programs, the Drosophila model organism Encyclopedia of DNA Elements (modENCODE) project is comprehensively mapping transcripts, histone modifications, chromosomal proteins, transcription factors, replication proteins and intermediates, and nucleosome properties across a developmental time course and in multiple cell lines. We have generated more than 700 data sets and discovered protein-coding, noncoding, RNA regulatory, replication, and chromatin elements, more than tripling the annotated portion of the Drosophila genome. Correlated activity patterns of these elements reveal a functional regulatory network, which predicts putative new functions for genes, reveals stage- and tissue-specific regulators, and enables gene-expression prediction. Our results provide a foundation for directed experimental and computational studies in Drosophila and related species and also a model for systematic data integration toward comprehensive genomic and functional annotation. Several years after the complete genetic sequencing of many species, it is still unclear how to translate genomic information into a functional map of cellular and developmental programs. The Encyclopedia of DNA Elements (ENCODE) (1) and model organism ENCODE (modENCODE) (2) projects use diverse genomic assays to comprehensively annotate the Homo sapiens (human), Drosophila melanogaster (fruit fly), and Caenorhabditis elegans (worm) genomes, through systematic generation and computational integration of functional genomic data sets. Previous genomic studies in flies have made seminal contributions to our understanding of basic biological mechanisms and genome functions, facilitated by genetic, experimental, computational, and manual annotation of the euchromatic and heterochromatic genome (3), small genome size, short life cycle, and a deep knowledge of development, gene function, and chromosome biology. The functions

A Multi-Hop Advertising Discovery and Delivering Protocol for Multi Administrative Domain MANET

Directory of Open Access Journals (Sweden)

Federico Mari

2013-01-01

Full Text Available A Mobile Ad-hoc NETwork (MANET is Multi Administrative Domain (MAD if each network node belongs to an independent authority, that is each node owns its resources and there is no central authority owning all network nodes. One of the main obstructions in designing Service Advertising, Discovery and Delivery (SADD protocol for MAD MANETs is the fact that, in an attempt to increase their own visibility, network nodes tend to flood the network with their advertisements. In this paper, we present a SADD protocol for MAD MANET, based on Bloom filters, that effectively prevents advertising floods due to such misbehaving nodes. Our results with the ns-2 simulator show that our SADD protocol is effective in counteracting advertising floods, it keeps low the collision rate as well as the energy consumption while ensuring that each peer receives all messages broadcasted by other peers.

RELAP5/MOD2 blind calculation of GERDA small break test and data comparison

International Nuclear Information System (INIS)

Ogden, D.M.; Steiner, J.L.; Waterman, M.E.

1985-01-01

The Idaho National Engineering Laboratory (INEL), in support of the USNRC, has developed a RELAP5/MOD2 model of the GERDA facility to be used for analysis of the GERDA data, particularly relative to the phenomena of natural circulation and the boiler condenser mode of heat transfer. A blind calculation of GERDA Test 1605AA and a preliminary comparison with experimental data has been performed. The GERDA facility is a single loop integral facility with an electrically heated core. A general arrangement diagram of the facility is shown. The GERDA facility was designed for the performance of both separate effects and overall systems tests

Current algorithms used in reactor safety codes and the impact of future computer development on these algorithms

International Nuclear Information System (INIS)

Mahaffy, J.H.; Liles, D.R.; Woodruff, S.B.

1985-01-01

Computational methods and solution procedures used in the US Nuclear Regulatory Commission's reactor safety systems codes, Transient Reactor Analysis Code (TRAC) and Reactor Leak and Power Safety Excursion Code (RELAP), are reviewed. Methods used in TRAC-PF1/MOD1, including the stability-enhancing two-step (SETS) technique, which permits fast computations by allowing time steps larger than the material Courant stability limit, are described in detail, and the differences from RELAP5/MOD2 are noted. Developments in computing, including parallel and vector processing, and their applicability to nuclear reactor safety codes are described. These developments, coupled with appropriate numerical methods, make detailed faster-than-real-time reactor safety analysis a realistic near-term possibility

Long Term Retention of Deuterium and Tritium in Alcator C-Mod

International Nuclear Information System (INIS)

FIORE, C.; LABOMBARD, B.; LIPSCHULTZ, B.; PITCHER, C.S.; SKINNER, C.H.; WAMPLER, WILLIAM R.

1999-01-01

We estimate the total in-vessel deuterium retention in Alcator C-Mod from a run campaign of about 1090 plasmas. The estimate is based on measurements of deuterium retained on 22 molybdenum tiles from the inner wall and divertor. The areal density of deuterium on the tiles was measured by nuclear reaction analysis. From these data, the in-vessel deuterium inventory is estimated to be about 0.1 gram, assuming the deuterium coverage is toroidally symmetric. Most of the retained deuterium is on the walls of the main plasma chamber, only about 2.5% of the deuterium is in the divertor. The D coverage is consistent with a layer saturated by implantation with ions and charge-exchange neutrals from the plasma. This contrasts with tokamaks with carbon plasma-facing components (PFC's) where long-term retention of tritium and deuterium is large and mainly in the divertor due to codeposition with carbon eroded by the plasma. The low deuterium retention in the C-Mod divertor is mainly due to the absence of carbon PFC's in C-Mod and the low erosion rate of Mo

Brazilian Irradiation Project: CAFE-MOD1 validation experimental program

International Nuclear Information System (INIS)

Mattos, Joao Roberto Loureiro de; Costa, Antonio Carlos L. da; Esteves, Fernando Avelar; Dias, Marcio Soares

1999-01-01

The Brazilian Irradiation Project whose purpose is to provide Brazil with a minimal structure to qualify the design, fabrication and quality procedures of nuclear fuels, consists of three main facilities: IEA-R1 reactor of IPEN-CNEN/SP, CAFE-MOD1 irradiation device and a unit of hot cells. The CAFE-MOD1 is based on concepts successfully used for more than 20 years in the main nuclear institutes around the world. Despite these concepts are already proved it should be adapted to each reactor condition. For this purpose, there is an ongoing experimental program aiming at the certification of the criteria and operational limits of the CAFE-MOD1 in order to get the allowance for its installation at the IEA-R1 reactor. (author)

THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

International Nuclear Information System (INIS)

Muramatsu, Ken; Akimoto, Masayuki

1982-08-01

THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

A modified MOD16 algorithm to estimate evapotranspiration over alpine meadow on the Tibetan Plateau, China

Science.gov (United States)

Chang, Y.; Ding, Y.; Zhao, Q.; Zhang, S.

2017-12-01

The accurate estimation of evapotranspiration (ET) is crucial for managing water resources in areas with extreme climates affected by climate change, such as the Tibetan Plateau (TP). The MOD16 ET product has also been validated and applied in many countries with various climates, however, its performance varies under different climates and regions. Several have studied ET based on satellite-based models on the TP. However, only a few studies on the performance of MOD16 in the TP with heterogeneous land cover have been reported. This study proposes an improved algorithm for estimating ET based on a proposed modified MOD16 method over alpine meadow on the TP in China. Wind speed and vegetation height were integrated to estimate aerodynamic resistance, while the temperature and moisture constraint for stomatal conductance were revised based on the technique proposed by Fisher et al. (2008). Moreover, Fisher's method for soil evaporation was introduced to decrease the uncertainty of soil evaporation estimation. Five representative alpine meadow sites on the TP were selected to investigate the performance of the modified algorithm. Comparisons between ET observed using Eddy Covariance (EC) and estimated using both the original method and modified method suggest that the modified algorithm had better performance than the original MOD16 method. This result was achieved considering that the coefficient of determination (R2) increased from 0.28 to 0.70, and the root mean square error (RMSE) decreased from 1.31 to 0.77 mm d-1. The modified algorithm also outperformed on precipitation days compared to non-precipitation days at Suli and Hulugou sites, while it performed well for both non-precipitation and precipitation days at Tanggula site. Comparisons of the 8-day ET estimation using the MOD16 product, original MOD16 method, and modified MOD16 method with observed ET suggest that MOD16 product underestimated ET over the alpine meadow of the TP during the growing season

RELAP5/MOD1-EUR evaluation. Comparison with the INEL original version

International Nuclear Information System (INIS)

Mazzantini, O.A.

1990-01-01

In this work, the values calculated from two versions of the RELAP5/MOD1 code are compared with those measured in different tests. The first version of RELAP5 is the cycle 19 of the original version of INEL (RELAP5/MOD1-INEL) and the second version improved by EURATOM (RELAP5/MOD1-EUR) which was transferred to ENACE through agreements made with SIEMENS/KWU. (Author) [es

BEACON/MOD3, 1-D and 2-D 2 Phase Flow and Heat Transfer in Containment, LWR LOCA

International Nuclear Information System (INIS)

Broadus, C.R.; Doyle, R.J.; James, S.W.; Lime, J.F.; Mings, W.J.; Ramsthaler, J.A.; Sahota, M.S.

1982-01-01

1 - Description of problem or function: The BEACON series of programs is designed to perform a best-estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped- parameter representations for the various parts of the system. BEACON/MOD3 contains mass and heat transfer models for wall film and for wall conduction, and is suitable for the evaluation of short- term transients in PWR dry containment systems. The capability to examine the details of a two-components, two-phase flow field in one or two dimensions under nonhomogeneous, nonequilibrium conditions (unequal velocities, unequal temperatures between the two phases) allows analysis of such problems as the calculation of jet impact forces of a fluid leaving a pipe break, the motion of a large pressure wave across a compartment, the variation in flow properties as air is displaced from a compartment by steam and water, the water entrainment or de-entrainment by a high-speed vapor flow, the flow of a flashing liquid, and many other complex nonequilibrium problems of containment system analyses. 2 - Method of solution: The basic Eulerian flow solution procedure is based on the K-FIX two-dimensional two-phase numerical method. Each phase is described by its own density, velocity, and temperature as determined by separate sets of mass, momentum, and energy equations. The two phases are coupled by exchange parameters which model the exchange of mass, momentum, and energy between the two phases. The two sets of field equations are solved with a Eulerian finite- difference technique that implicitly treats the phase transitions and inter-phasic heat transfer in the pressure iteration. The implicit solution is accomplished iteratively without linearization and allows both phases to be

Nd3-xBixFe4GaO12 (x = 2, 2.5 films on glass substrates prepared by MOD method

Directory of Open Access Journals (Sweden)

Yoshida T.

2014-07-01

Full Text Available We studied Nd3-XBiXFe4GaO12 films to obtain perpendicular magnetic anisotropy as well as large Faraday effect. NdBi2Fe4GaO12 (Bi2:NIGG and Nd0.5Bi2.5Fe4GaO12 (Bi2.5:NIGG films were obtained on Nd2BiFe4GaO12 (Bi1:NIGG layer prepared on glass substrates by metal-organic decomposition (MOD method. Bi2:NIGG and Bi2.5:NIGG films showed large Faraday rotation angles of 7.5 and 10.5 degree/µm, at a wavelength of 520 nm, respectively. Those films have perpendicular magnetic anisotropy with a coercivity of 350 Oe and a saturation magnetic field of 730 Oe.

REACT-Mod: a mathematical model for transient calculation of chemical reactions with U-Pu-Np-Tc in the aqueous nitric acid solution

International Nuclear Information System (INIS)

Tachimori, Shoichi; Kitamura, Tatsuaki.

1996-10-01

A computer code REACT-Mod which simulates various chemical reactions in an aqueous nitric acid solution involving uranium, plutonium, neptunium, technetium etc. e.g., redox, radiolytic and disproportionation reactions of 68, was developed based on the kinetics model. The numerical solution method adopted in the code are two, a kinetics model totally based on the rate law of which differential equations are solved by the modified Porsing method, and a two-step model based on both the rate law and equilibrium law. Only the former treats 27 radiolytic reactions. The latter is beneficially used to have a quick and approximate result by economical computation. The present report aims not only to explain the concept, chemical reactions treated and characteristics of the model but also to provide details of the program for users of the REACT-Mod code. (author)

Sur la modélisation des supraconducteurs : le ``modèle de l'état critique'' de Bean, en trois dimensions

Science.gov (United States)

Bossavit, A.

1993-03-01

Macroscopic modelling of superconductors demands a substitution of some nonlinear behavior law for Ohm's law. For this, a version of Bean's “critical state” model, derived from the setting of a convex functional of the current density field, valid in dimension 3 without any previous assumption about the direction of currents, is proposed. It is shown how two standard three-dimensional finite element methods (“h-formulation” and “e-formulation”), once fitted with this model, can deal with situations were superconductors are present. La modélisation macroscopique des supraconducteurs suppose le remplacement de la loi d'Ohm par une loi de comportement non linéaire adéquate. On présente à cet effet une version du “modèle de Bean”, ou modèle de l'état critique, basée sur la construction d'une certaine fonctionnelle convexe du champ des densités de courant, qui est valable en dimension 3 sans hypothèses préalables sur la direction des courants. On montre comment adapter deux méthodes standards de calcul de courants de Foucault par élérnents finis en trois dimensions (“en h” et “en e”) à la présence de supraconducteurs, en incorporant ce modèle.

Plans and status of RELAP5/MOD3

International Nuclear Information System (INIS)

Weaver, W.L.

1989-01-01

RELAP5/MOD3 is a pressurized water reactor (PWR) system analysis code being developed jointly by the US Nuclear Regulatory Commission (USNRC) and consisting of several of the countries that are members of the International Code Assessment and Applications Program (ICAP). This code development program is called the ICAP Code Improvement Program. The mission of the RELAP5/MOD3 code improvement program is to develop a code version suitable for the analysis of all transients and postulated accidents in PER systems including both large and small break loss of coolant accidents (LOCA's) as well as the full range of operational transients. The emphasis of the RELAP5/MOD3 development will be on large break LOCA since previous versions of RELAP5 were developed for and assessed against small break LOCA and operation transient test data. The paper discusses the various code models to be improved and presents the results of work completed to date

Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

International Nuclear Information System (INIS)

Bujan, A.; Adamik, V.; Misak, J.

1986-01-01

A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

Assessment of RELAP/MOD3 using BETHSY 6.2TC 6-inch cold leg side break comparative test

International Nuclear Information System (INIS)

Chung, Young-Jong; Jeong, Jae-Jun; Chang, Won-Pyo; Kim, Dong-Su

1996-10-01

This report presents the results of the RELAP5/MOD3 Version 7j assessment on BETHSY 6.2TC. BETHSY 6.2TC test corresponding to a six inch cold leg break LOCA of the Pressurizer Water Reactor(PWR). The primary objective of the test was to provide reference data of two facilities of different scales (BETHSY and LSTF facility). On the other hand, the present calculation aims at analysis of RELAP5/N4OD3 capability on the small break LOCA simulation, The results of calculation have shown that the RELAP5/MOD3 reasonably predicts occurrences as well as trends of the major phenomena such as primary pressure, timing of loop seal clearing, liquid hold up, etc. However, some disagreements also have been found in the predictions of loop seal clearing, collapsed core water level after loop seal clearing, and accumulator injection behaviors. For better understanding of discrepancies in same predictions, several sensitivity calculations have been performed as well. These include the changes of two-phase discharge coefficient at the break junction and some corrections of the interphase drag term. As result, change of a single parameter has not improved the overall predictions and it has been found that the interphase drag model has still large uncertainties

Les enjeux du modèle standard

CERN Document Server

CERN. Geneva

2008-01-01

Toute la matière visible dans l'Univers est décrite par le « Modèle Standard ». Selon cette théorie, la matière est constituée d'atomes, qui contiennent des électrons orbitant autour de noyaux, dont les composants fondamentaux sont les quarks. Quatre forces fondamentales agissent entre ces particules élémentaires : les forces électromagnétique est gravitationnelle, et les interactions nucléaires forte et faible. La description fournie par le Modèle Standard de ces particules et de leurs interactions est en parfait accord avec les expériences. Néanmoins, des questions fondamentales restent sans réponse jusqu'à maintenant : d'où vient la masse des particules ? Pourquoi y a-t-il tant de types de particules ? Existe-t-il une théorie unifiée de toutes les interactions ? Quelle est la nature de la matière cachée prédite par les astrophysiciens ? Le LHC au CERN donnera des éléments de réponse à ces questions au-delà du Modèle Standard.

It's Time to Take a Stand.

Science.gov (United States)

Wallace, Stephen G.

1987-01-01

Counselors and campers at Cape Cod Sea Camps in Brewster, Massachusetts, banded together to combat the problem of drinking and driving by forming the first camp-based chapter of Students Against Driving Drunk (SADD). The chapter's success shows that SADD can be an asset to camp programs. (JHZ)

Control of substrate oxidation in MOD ceramic coating on low-activation ferritic steel with reduced-pressure atmosphere

Energy Technology Data Exchange (ETDEWEB)

Tanaka, Teruya, E-mail: teru@nifs.ac.jp; Muroga, Takeo

2014-12-15

Highlights: • A Cr{sub 2}O{sub 3} layer was produced on a ferritic steel substrate with a reduced-pressure. • The Cr{sub 2}O{sub 3} layer prevents further substrate oxidation in following coating process. • The Cr{sub 2}O{sub 3} layer has a function as a hydrogen permeation barrier. • A smooth MOD Er{sub 2}O{sub 3} coating was successfully made on the Cr{sub 2}O{sub 3} layer by dip coating. • The Cr{sub 2}O{sub 3} layer would enhance flexibility in MOD coating process and performances. - Abstract: An Er{sub 2}O{sub 3} ceramic coating fabricated using the metal–organic decomposition (MOD) method on a Cr{sub 2}O{sub 3}-covered low-activation ferritic steel JLF-1 substrate was examined to improve hydrogen permeation barrier performance of the coating. The Cr{sub 2}O{sub 3} layer was obtained before coating by heat treating the substrate at 700 °C under reduced pressures of <5 × 10{sup −3} Pa and 5 Pa. The Cr{sub 2}O{sub 3} layer was significantly stable even with heat treatment at 700 °C in air. This layer prevented further production of Fe{sub 2}O{sub 3}, which has been considered to degrade coating performance. An MOD Er{sub 2}O{sub 3} coating with a smooth surface was successfully obtained on a Cr{sub 2}O{sub 3}-covered JLF-1 substrate by dip coating followed by drying and baking. Preprocessing to obtain a Cr{sub 2}O{sub 3} layer would provide flexibility in the coating process for blanket components and ducts. Moreover, the Cr{sub 2}O{sub 3} layer suppressed hydrogen permeation through the JLF-1 substrate. While further optimization of the coating fabrication process is required, it would be possible to suppress hydrogen permeation significantly by multilayers of Cr{sub 2}O{sub 3} and MOD oxide ceramic.

Modèle exploitable pour la définition de la commande du robot

OpenAIRE

ZIMMER-CHEVRET, Sandra; LANGLOIS, Laurent; BEN ATTAR, Amarilys

2014-01-01

Ce document traite de la modélisation des actions mécaniques entre l’outil et la matière. L’objectif est de définir un modèle exploitable pour la définition de la commande du robot. Dans un premier temps, le rapport présente une synthèse bibliographique des modèles des interactions mécaniques développés à ce jour. Pour un modèle choisi, les paramètres constituant ce dernier ont été calculés à partir de données expérimentales. Puis, la validité du modèle a été étudiée. Pour une même configurat...

Analysis of C-MOD molybdenum divertor erosion and code/data comparison

Energy Technology Data Exchange (ETDEWEB)

Brooks, J.N., E-mail: brooksjn@purdue.edu [Purdue University, West Lafayette, IN (United States); Allain, J.P. [Purdue University, West Lafayette, IN (United States); Whyte, D.G.; Ochoukov, R.; Lipschultz, B. [Massachusetts Institute of Technology, Cambridge, MA (United States)

2011-08-01

We analyze an important 15 year old Alcator C-MOD study of campaign-integrated molybdenum divertor erosion in which the measured net erosion was significantly higher ({approx}X3) than originally predicted by a simple model . We perform full process sputtering erosion/redeposition computational analysis including the effect of a possible RF induced sheath. The simulations show that most sputtered Mo atoms are ionized close to the surface and strongly redeposited, via Lorentz force motion and collisional friction with the high density incoming plasma. The predicted gross erosion profile is a reasonable match to MoI influx data, however, the critically important net erosion comparison with post-exposure Mo tile analysis is poor, with {approx}X10 higher peak erosion measured than computed. An RF sheath increases predicted erosion by {approx}45%, thus being significant but not fundamental. We plan future analysis.

Molybdate transporter ModABC is important for Pseudomonas aeruginosa chronic lung infection.

Science.gov (United States)

Périnet, Simone; Jeukens, Julie; Kukavica-Ibrulj, Irena; Ouellet, Myriam M; Charette, Steve J; Levesque, Roger C

2016-01-12

Mechanisms underlying the success of Pseudomonas aeruginosa in chronic lung infection among cystic fibrosis (CF) patients are poorly defined. The modA gene was previously linked to in vivo competitiveness of P. aeruginosa by a genetic screening in the rat lung. This gene encodes a subunit of transporter ModABC, which is responsible for extracellular uptake of molybdate. This compound is essential for molybdoenzymes, including nitrate reductases. Since anaerobic growth conditions are known to occur during CF chronic lung infection, inactivation of a molybdate transporter could inhibit proliferation through the inactivation of denitrification enzymes. Hence, we performed phenotypic characterization of a modA mutant strain obtained by signature-tagged mutagenesis (STM_modA) and assessed its virulence in vivo with two host models. The STM_modA mutant was in fact defective for anaerobic growth and unable to use nitrates in the growth medium for anaerobic respiration. Bacterial growth and nitrate usage were restored when the medium was supplemented with molybdate. Most significantly, the mutant strain showed reduced virulence compared to wild-type strain PAO1 according to a competitive index in the rat model of chronic lung infection and a predation assay with Dictyostelium discoideum amoebae. As the latter took place in aerobic conditions, the in vivo impact of the mutation in modA appears to extend beyond its effect on anaerobic growth. These results support the modABC-encoded transporter as important for the pathogenesis of P. aeruginosa, and suggest that enzymatic machinery implicated in anaerobic growth during chronic lung infection in CF merits further investigation as a potential target for therapeutic intervention.

KONVENCIJE MOD O PRISILNEM DELU V LUČI DANAŠNJEGA ČASA

OpenAIRE

Kokoravec, Mateja

2015-01-01

Prisilno delo je antiteza dostojnega dela, za katerega se zavzema MOD. Za 21 milijonov ljudi to ni samo bled spomin iz preteklosti, ampak težka realnost še danes. Mednarodni delovni standardi, ki jih oblikuje MOD, predstavljajo temeljne minimalne socialne standarde, dogovorjene s strani glavnih udeležencev v svetovnem gospodarstvu. Z njimi si MOD prizadeva doseči dostojne delovne pogoje, saj je v današnjem svetu globalizacije za doseganje teh ciljev potrebno ukrepanje na mednarodni ravni. ...

The mod industries? The industrial logic of non-market game production

NARCIS (Netherlands)

Nieborg, D.B.; van der Graaf, S.

2008-01-01

This article seeks to make the relationship between non-market game developers (modders) and the game developer company explicit through game technology. It investigates a particular type of modding, i.e. total conversion mod teams, whose organization can be said to conform to the high-risk,

Latest developments for a computer aided thermohydraulic network

International Nuclear Information System (INIS)

Alemberti, A.; Graziosi, G.; Mini, G.; Susco, M.

1999-01-01

Thermohydraulic networks are I-D systems characterized by a small number of basic components (pumps, valves, heat exchangers, etc) connected by pipes and limited spatially by a defined number of boundary conditions (tanks, atmosphere, etc). The network system is simulated by the well known computer program RELAPS/mod3. Information concerning the network geometry component behaviour, initial and boundary conditions are usually supplied to the RELAPS code using an ASCII input file by means of 'input cards'. CATNET (Computer Aided Thermalhydraulic NETwork) is a graphically user interface that, under specific user guidelines which completely define its range of applicability, permits a very high level of standardization and simplification of the RELAPS/mod3 input deck development process as well as of the output processing. The characteristics of the components (pipes, valves, pumps etc), defining the network system can be entered through CATNET. The CATNET interface is provided by special functions to compute form losses in the most typical bending and branching configurations. When the input of all system components is ready, CATNET is able to generate the RELAPS/mod3 input file. Finally, by means of CATNET, the RELAPS/mod3 code can be run and its output results can be transformed to an intuitive display form. The paper presents an example of application of the CATNET interface as well as the latest developments which greatly simplified the work of the users and allowed to reduce the possibility of input errors. (authors)

AP600 large-break loss-of-collant-accident developmental assessment plan for TRAC-PF1/MOD2

International Nuclear Information System (INIS)

Knight, T.D.

1996-07-01

The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems to protect the plant against possible accidents and transients. The design has been submitted to the U.S. NRC for design certification. The NRC has selected the Transient Reactor Analysis Code (TRAC)-PF1/MOD2 for performing large break loss-of coolant-accident (LBLOCA) analysis to support the certification effort. This document defines the tests to be used in the current phase of developmental assessment related to AP600 LBLOCA

Development of sorbent therapy for multiple organ dysfunction syndrome (MODS)

International Nuclear Information System (INIS)

Li Li; Pan Jilun; Yu Yaoting

2007-01-01

As a syndrome, multiple organ dysfunction (MODS) is defined as an altered organ function in the setting of sepsis, septic shock or systemic inflammatory response syndrome (SIRS) and is the most common cause of death in intensive care units. Endotoxin, a constituent of cell walls of Gram-negative bacteria, plays an important role in the initiation and development of MODS. The cytokines, especially tumor necrosis factor alpha (TNF-alpha), are early regulators of the immune response and can induce the release of secondary cytokines. To remove endotoxin and TNF-alpha from patients with MODS, the adsorption method has proven to be most effective. In this review, we provide various methods of removal of endotoxins and TNF-alpha using different adsorbents. (topical review)

Flux-driven turbulence GDB simulations of the IWL Alcator C-Mod L-mode edge compared with experiment

Science.gov (United States)

Francisquez, Manaure; Zhu, Ben; Rogers, Barrett

2017-10-01

Prior to predicting confinement regime transitions in tokamaks one may need an accurate description of L-mode profiles and turbulence properties. These features determine the heat-flux width upon which wall integrity depends, a topic of major interest for research aid to ITER. To this end our work uses the GDB model to simulate the Alcator C-Mod edge and contributes support for its use in studying critical edge phenomena in current and future tokamaks. We carried out 3D electromagnetic flux-driven two-fluid turbulence simulations of inner wall limited (IWL) C-Mod shots spanning closed and open flux surfaces. These simulations are compared with gas puff imaging (GPI) and mirror Langmuir probe (MLP) data, examining global features and statistical properties of turbulent dynamics. GDB reproduces important qualitative aspects of the C-Mod edge regarding global density and temperature profiles, within reasonable margins, and though the turbulence statistics of the simulated turbulence follow similar quantitative trends questions remain about the code's difficulty in exactly predicting quantities like the autocorrelation time A proposed breakpoint in the near SOL pressure and the posited separation between drift and ballooning dynamics it represents are examined This work was supported by DOE-SC-0010508. This research used resources of the National Energy Research Scientific Computing Center (NERSC).

Measurement of local critical currents in TFA-MOD processed coated conductors by use of scanning Hall-probe microscopy

International Nuclear Information System (INIS)

Shiohara, K.; Higashikawa, K.; Kawaguchi, T.; Inoue, M.; Kiss, T.; Yoshizumi, M.; Izumi, T.

2011-01-01

We have investigated 2-dimensional distribution of critical current density. We have measured TFA-MOD processed YBCO coated conductor. We used scanning Hall-probe microscopy. These provided information is useful for fabrication process of coated conductor. We have carried out 2-dimensional (2D) measurement of local critical current in a Trifluoroacetates-Metal Organic Deposition (TFA-MOD) processed YBCO coated conductor using scanning Hall-probe microscopy. Recently, remarkable R and D accomplishments on the fabrication processes of coated conductors have been conducted extensively and reported. The TFA-MOD process has been expected as an attractive process to produce coated conductors with high performance at a low production cost due to a simple process using non-vacuum equipments. On the other hand, enhancement of critical currents and homogenization of the critical current distribution in the coated conductors are definitely very important for practical applications. According to our measurements, we can detect positions and spatial distribution of defects in the conductor. This kind of information will be very helpful for the improvement of the TFA-MOD process and for the design of the conductor intended for practical electric power device applications.

Alcator C-MOD final safety analysis

International Nuclear Information System (INIS)

Fiore, C.L.

1989-06-01

This document is designed to address the safety issues involved with the Alcator C-Mod project. This report will begin with a brief description of the experimental objectives which will be followed by information concerning the site. The Alcator C-Mod experiment is a pulsed fusion experiment in which a plasma formed from small amounts of hydrogen or deuterium gas is confined in a magnetic field for short periods (∼1 s). No radioactive fuels or fissile materials are used in the device, so that no criticality hazard exists and no credible nuclear accident can occur. During deuterium operation, the production of a small number of neutrons from a short pulse could result in a small amount of short- and intermediate-lived radioactive isotopes being produced inside the experimental cell. This report will demonstrate that this does not pose an additional hazard to the general population. The health and safety hazards resulting from Alcator C-Mod occur to the workers on the experiment, each of which is described in its own chapter with the steps taken to minimize the risk to employees. These hazards include fire, chemicals and cryogenics, air quality, electrical, electromagnetic radiation, ionizing radiation, and mechanical and natural phenomena. None of these hazards is unique to the facility, and methods of protection from them are well defined and are discussed in the chapter which describes each hazard. The quality assurance program, critical to ensuring the safety aspects of the program, will also be described

Kyllinger har effektivt immunforsvar mod herpes

DEFF Research Database (Denmark)

Buchmann, Kurt

2008-01-01

Forskere ved Københavns Universitet, Fakultet for Biovidenskab har studeret kyllingens MHC molekyler og kan derigennem forklare hvorfor en særlig stamme indenfor denne art er modstandsdygtige mod en særlig herpesvirus. Udgivelsesdato: 18. januar 2008...

Estimation des paramètres de modèles pour la digestion anaérobie

OpenAIRE

Martinez , Ricardo

2003-01-01

La finalité de ce travail a été d'estimer de la meilleure manière possible des paramètres de différents modèles sur la respiration anaérobie. Dans un premier temps nous présentons les bases qu'on utilise pour construire les modèles du procédé biochimique. Puis nous faisons une analyse qualitative des modèles les plus connus sur le sujet. Dans un deuxième temps nous avons suggéré différents modèles, afin de mieux estimer les paramètres du modèle initial. Nous avons mis au point un programme d'...

The LHCD Launcher for Alcator C-Mod - Design, Construction, Calibration and Testing

International Nuclear Information System (INIS)

Hosea, J.; Beals, D.; Beck, W.; Bernabei, S.; Burke, W.; Childs, R.; Ellis, R.; Fredd, E.; Greenough, N.; Grimes, M.; Gwinn, D.; Irby, J.; Jurczynski, S.; Koert, P.; Kung, C.C.; Loesser, G.D.; Marmar, E.; Parker, R.; Rushinski, J.; Schilling, G.; Terry, D.; Vieira, R.; Wilson, J.R.; Zaks, J.

2005-01-01

MIT and PPPL have joined together to fabricate a high-power lower hybrid current drive (LHCD) system for supporting steady-state AT regime research on Alcator C-Mod. The goal of the first step of this project is to provide 1.5 MW of 4.6 GHz rf [radio frequency] power to the plasma with a compact launcher which has excellent spectral selectivity and fits into a single C-Mod port. Some of the important design, construction, calibration and testing considerations for the launcher leading up to its installation on C-Mod are presented here

Assessment of RELAP5 MOD3.3 and CATHARE 2 V1.5A against a full scale test of PERSEO device

International Nuclear Information System (INIS)

Bianchi, F.; Meloni, P.; Ferri, R.; Achilli, A.

2004-01-01

PERSEO device was developed in the framework of a domestic research program on innovative safety systems, with the purpose to increase the reliability of passive Decay Heat Removal Systems implementing in-pool heat exchangers. The device was tested at SIET Thermal-hydraulic Research Centre by modifying the existing PANTHERS IC-PCC facility. Two types of tests were performed: integral tests and stability tests. The experimental data acquired in the test campaign allowed a validation of a RELAP5/mod 3.3 beta release and CATHARE2 V1.5a/Mod8.1 full scale model of the PERSEO device. The paper deals with the comparison between the two codes against an integral test considered representative from the point of view of the PERSEO functioning and it highlights capabilities and limits of the codes in simulating such kind of test. (authors)

Mod-5A Wind Turbine Generator Program Design Report. Volume 2: Conceptual and Preliminary Design, Book 1

Science.gov (United States)

1984-01-01

The design, development and analysis of the 7.3 MW MOD-5A wind turbine generator is documented. There are four volumes. In Volume 2, book 1 the requirements and criteria for the design are presented. The conceptual design studies, which defined a baseline configuration and determined the weights, costs and sizes of each subsystem, are described. The development and optimization of the wind turbine generator are presented through the description of the ten intermediate configurations between the conceptual and final designs. Analyses of the system's load and dynamics are presented.

Fatigue Properties of Aged Mod. 9Cr-1Mo

International Nuclear Information System (INIS)

Kim, Dae Whan; Kim, Sung Ho; Lee, Chan Bock

2007-01-01

Ferritic/Martensitic steel has a good mechanical properties and a lower thermal expansion coefficient than austenitic stainless steel. Mechanical property of Mod. 9Cr-1Mo steel is less than austenitic stainless steel at high temperature. High temperature mechanical properties are affected by precipitation for Mod. 9Cr-1Mo. FMS steel is used for long time at high temperature and the effect of aging on mechanical properties is very important. In this study, low cycle fatigue properties with aging were investigated

Model of nuclear reactor type VVER-1000/V-320 built by computer code ATHLET-CD

International Nuclear Information System (INIS)

Georgiev, Yoto; Filipov, Kalin; Velev, Vladimir

2014-01-01

A model of nuclear reactor type VVER-1000 V-320 developed for computer code ATHLET-CD2.1A is presented. Validation of the has been made, in the analysis of the station blackout scenario with LOCA on fourth cold leg is shown. As the calculation has been completed, the results are checked through comparison with the results from the computer codes ATHLET-2.1A, ASTEC-2.1 and RELAP5mod3.2

Vold mod børn

DEFF Research Database (Denmark)

Christensen, Else; Agerlund Sloth Larsen, Dorthe

er der gennemført en interviewundersøgelse, hvor i alt 14 sagsbehandlere fra fire forskellige store kommuner er interviewet om deres erfaringer fra sager med (mistanke om) fysisk vold mod børn, om hvordan sådanne sager sædvanligvis starter i socialforvaltningen, om undersøgelsesforløbet, om...

Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

Energy Technology Data Exchange (ETDEWEB)

Banati, J. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

1995-09-01

This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

Three-dimensional equilibria and transport in RFX-mod: A description using stellarator tools

International Nuclear Information System (INIS)

Gobbin, M.; Bonfiglio, D.; Lorenzini, R.; Marrelli, L.; Martin, P.; Martines, E.; Momo, B.; Predebon, I.; Puiatti, M. E.; Spizzo, G.; Terranova, D.; Boozer, A. H.; Cooper, A. W.; Escande, D. F.; Hirshman, S. P.; Lore, J.; Sanchez, R.; Spong, D. A.; Pomphrey, N.

2011-01-01

RFX-mod self-organized single helical axis (SHAx) states provide a unique opportunity to advance 3D fusion physics and establish a common knowledge basis in a parameter region not covered by stellarators and tokamaks. The VMEC code has been adapted to the reversed-field pinch (RFP) to model SHAx equilibria in fixed boundary mode with experimental measurements as constraint. The averaged particle diffusivity over the helical volume, estimated with the Monte Carlo code ORBIT, has a neoclassical-like dependence on collisionality and does not show the 1/ν trend of un-optimized stellarators. In particular, the helical region boundary, corresponding to an electron transport barrier with zero magnetic shear and improved confinement, has been investigated using numerical codes common to the stellarator community. In fact, the DKES/PENTA codes have been applied to RFP for local neoclassical transport computations, including radial electric field, to estimate thermal diffusion coefficients in the barrier region for typical RFX-mod temperature and density profiles. A comparison with power balance estimates shows that residual chaos due to secondary tearing modes and small-scale turbulence still contribute to drive anomalous transport in the barrier region.

Identification of Bacillus thuringiensis Cry1AbMod binding-proteins from Spodoptera frugiperda.

Science.gov (United States)

Martínez de Castro, Diana L; García-Gómez, Blanca I; Gómez, Isabel; Bravo, Alejandra; Soberón, Mario

2017-12-01

Bacillus thuringiensis Cry toxins are currently used for pest control in transgenic crops but evolution of resistance by the insect pests threatens the use of this technology. The Cry1AbMod toxin was engineered to lack the alpha helix-1 of the parental Cry1Ab toxin and was shown to counter resistance to Cry1Ab and Cry1Ac toxins in different insect species including the fall armyworm Spodoptera frugiperda. In addition, Cry1AbMod showed enhanced toxicity to Cry1Ab-susceptible S. frugiperda populations. To gain insights into the mechanisms of this Cry1AbMod-enhanced toxicity, we isolated the Cry1AbMod toxin binding proteins from S. frugiperda brush border membrane vesicles (BBMV), which were identified by pull-down assay and liquid chromatography-tandem mass spectrometry (LC-MS/MS). The LC-MS/MS results indicated that Cry1AbMod toxin could bind to four classes of aminopeptidase (N1, N3, N4 y N5) and actin, with the highest amino acid sequence coverage acquired for APN 1 and APN4. In addition to these proteins, we found other proteins not previously described as Cry toxin binding proteins. This is the first report that suggests the interaction between Cry1AbMod and APN in S. frugiperda. Copyright © 2017 Elsevier Inc. All rights reserved.

Modernization and restructuring of realistic thermal hydraulic system analysis code, RELAP5/MOD3.3.1.2

International Nuclear Information System (INIS)

Chung, Bub Dong; Lee, Won Jae; Jeong, Jae Jun; Ha, Kwi Seok

1998-01-01

The code architecture entails the programming language and the code database. Various recent programming languages such as C, C ++ , Fortran 90, were considered as the candidate language for the modernization of RELAP5/MOD3.2.1.2. Among them, Fortran 90 was selected as a basic programming laguage for the modernization and restructuring of the code. Most of header file ( * .h) and equivalenced variables in RELAP5 have been replaced with members in the MODULE, which greatly enhance the code maintenance and readability. The FTB package is used for the dynamic memory management (DMM) of RELAP5. Although FTB DMM features are very successful, the use of FTB has been the obstacle in the maintenance of the code. It is difficult to understand and change the coding, and it requires a significant effort to find out index errors in large memory pools. With new features introduced in Fortran 90, it is possible to slove dynamic allocation problems within the standard features in an elegant, clear safe way. Each of FTB data blocks can be replaced by the suitably organized derived variables in MODULE and the standard DMM scheme. This DMM scheme provides the code flexibility which can save the memory requirements depending on the problem sizes without a extensive use of the complex FTB package. The current user's interface of the RELAP5 consists of a set of input file, output file, and restart/plot file. Many users complain that this interface is not user friendly. It was mainly caused by the text-oriented programming, namly console programming during the past many years. Now, windows programming has become popular in most areas of software development. Using this windows programming technique, the user friend freatures can be implemented. The Visual Fortran Quick Win run-time library helps to turn graphics programs into simple Windows applications. RELAP5 code has been re-compiled with the Quick Win feature, and the mask for user's dialog and graphical x-y plot were designed. This

The mod industries? The industrial logic of non-market game production

OpenAIRE

2008-01-01

Abstract This article seeks to make the relationship between non-market game developers (modders) and the game developer company explicit through game technology. It investigates a particular type of modding, i.e. total conversion mod teams, whose organization can be said to conform to the high-risk, technologically-advanced, capital-intensive, proprietary practice of the developer company. The notion ...

Recueil de modèles aléatoires

CERN Document Server

Chafai, Djalil

2016-01-01

Ce recueil puise sa source dans les cours de master de mathématiques appliquées et de préparation à l’épreuve de modélisation de l’agrégation de mathématiques. Le parti pris de cet ouvrage est de polariser la rédaction par les modèles plutôt que par les outils, et de consacrer chaque chapitre à un modèle. Le premier public visé est celui des enseignants-chercheurs en probabilités, débutants ou confirmés. De nombreux chapitres peuvent également bénéficier directement à des étudiants de master ou préparant l’agrégation. Collected Stochastic Models This collection was inspired by applied mathematics Master classes in stochastic modeling. The focus is on models rather than on tools, and each chapter is devoted to a specific model. Though the book is primarily intended for academics in the field of probability theory, beginners and experienced researchers alike, many chapters will also benefit students preparing to pursue their Master degree in mathematics. .

The assessment of RELAP5/MOD2 based on pressurizer transient experiments

International Nuclear Information System (INIS)

Xue Hanjun; Tanrikut, A.; Menzel, R.

1992-03-01

Two typical experiments have been performed in Chinese test facility under full pressure load corresponding to typical PWRs, 1) dynamic behavior of pressurizer due to relief valve operations (Case-I) is extremely important in transients and accident conditions regarding depressurization of PWR primary system; 2) Outsurge/Insurge operation is one of the transient which is often encountered and experienced in pressurizer systems due to pressure transients in primary system of PWRs. The simulation capability of RELAP5/MOD2 is good in comparison to experimental results. The physical models (such as interface model, stratification model), playing a major role in such simulation, seems to be realistic. The effect of realistic valve modeling in depressurization simulation is extremely important. Sufficient data for relief valve (the dynamic characteristics of valve) play a major role. The time dependent junction model and the trip valve model with a reduced discharge coefficient of 0.2 give better predictions in agreement with the experiment data while the trip valve models with discharge coefficient 1.0 yield overdepressurization. The simulation of outsurge/insurge transient yields satisfactory results. The thermal non-equilibrium model is important with respect to simulation of complicated physical phenomena in outsurge/insurge transient but has a negligible effect upon the depressurization simulation. (orig./HP)

The molybdate-binding protein (ModA) of the plant pathogen Xanthomonas axonopodis pv. citri.

Science.gov (United States)

Balan, Andrea; Santacruz, Carolina P; Moutran, Alexandre; Ferreira, Rita C C; Medrano, Francisco J; Pérez, Carlos A; Ramos, Carlos H I; Ferreira, Luís C S

2006-12-01

The modABC operon of phytopathogen Xanthomonas axonopodis pv. citri (X. citri) encodes a putative ABC transporter involved in the uptake of the molybdate and tungstate anions. Sequence analyses showed high similarity values of ModA orthologs found in X. campestris pv. campestris (X. campestris) and Escherichia coli. The X. citri modA gene was cloned in pET28a and the recombinant protein, expressed in the E. coli BL21 (DE3) strain, purified by immobilized metal affinity chromatography. The purified protein remained soluble and specifically bound molybdate and tungstate with K(d) 0.29+/-0.12 microM and 0.58+/-0.14 microM, respectively. Additionally binding of molybdate drastically enhanced the thermal stability of the recombinant ModA as compared to the apoprotein. This is the first characterization of a ModA ortholog expressed by a phytopathogen and represents an important tool for functional, biochemical and structural analyses of molybdate transport in Xanthomonas species.

Moving Past "Hello World": Learning to Mod in an Online Affinity Space

Science.gov (United States)

Subramanian, Shree Durga

2012-01-01

Game modding has increasingly become a mainstream and "cutting edge" medium to foster a broad range of critical software design and programming practices to learners coming from wide-ranging educational and professional backgrounds. Participatory practices, like game modding, are highly interest-driven and entail intense engagement with…

MOD approach for the growth of epitaxial CeO2 buffer layers on biaxially textured Ni-W substrates for YBCO coated conductors

International Nuclear Information System (INIS)

Bhuiyan, M S; Paranthaman, M; Sathyamurthy, S; Aytug, T; Kang, S; Lee, D F; Goyal, A; Payzant, E A; Salama, K

2003-01-01

We have grown epitaxial CeO 2 buffer layers on biaxially textured Ni-W substrates for YBCO coated conductors using a newly developed metal organic decomposition (MOD) approach. Precursor solution of 0.25 M concentration was spin coated on short samples of Ni-3 at%W (Ni-W) substrates and heat-treated at 1100 C in a gas mixture of Ar-4%H 2 for 15 min. Detailed x-ray studies indicate that CeO 2 films have good out-of-plane and in-plane textures with full-width-half-maximum values of 5.8 deg. and 7.5 deg., respectively. High temperature in situ XRD studies show that the nucleation of CeO 2 films starts at 600 C and the growth completes within 5 min when heated at 1100 C. SEM and AFM investigations of CeO 2 films reveal a fairly dense microstructure without cracks and porosity. Highly textured YSZ barrier layers and CeO 2 cap layers were deposited on MOD CeO 2 -buffered Ni-W substrates using rf-magnetron sputtering. Pulsed laser deposition (PLD) was used to grow YBCO films on these substrates. A critical current, J c , of about 1.5 MA cm -2 at 77 K and self-field was obtained on YBCO (PLD)/CeO 2 (sputtered)/YSZ (sputtered)/CeO 2 (spin-coated)/Ni-W

Evaluation of piping heat transfer, piping flow regimes, and steam generator heat transfer for the Semiscale Mod-1 isothermal tests

International Nuclear Information System (INIS)

French, R.T.

1975-08-01

Selected experimental data pertinent to piping heat transfer, transient fluid flow regimes, and steam generator heat transfer obtained during the Semiscale Mod-1 isothermal blowdown test series (Test Series 1) are analyzed. The tests in this first test series were designed to provide counterparts to the LOFT nonnuclear experiments. The data from the Semiscale Mod-1 intact and broken loop piping are evaluated to determine the surface heat flux and average heat transfer coefficients effective during the blowdown transient and compared with well known heat transfer correlations used in the RELAP4 computer program. Flow regimes in horizontal pipe sections are calculated and compared with data obtained from horizontal and vertical densitometers and with an existing steady state flow map. Effects of steam generator heat transfer are evaluated quantitatively and qualitatively. The Semiscale Mod-1 data and the analysis presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict system response to piping heat transfer, piping flow regimes, and steam generator heat transfer during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). 16 references. (auth)

Enhancement of flux pinning of TFA-MOD YBCO thin films by embedded nanoscale Y2O3

International Nuclear Information System (INIS)

Cui, X M; Tao, B W; Tian, Z; Xiong, J; Zhang, X F; Li, Y R

2006-01-01

YBCO films with different levels of excess yttrium were prepared on single-crystal LaAlO 3 with metal-organic deposition using trifluoroacetates (TFA-MOD). X-ray diffraction and transmission electron microscope measurements revealed excess yttrium in YBCO in the form of nanoscale Y 2 O 3 with (400) preferred orientation. The field dependence of J c demonstrated that YBCO films with Y 2 O 3 doping had enhanced J c in comparison with stoichiometric YBCO films in the magnetic fields. We think the reason for this is that the Y 2 O 3 nanoparticles act as pinning centres. YBCO films with 60% yttrium excess display 43% increased J c compared to stoichiometric YBCO films at a magnetic field of 1 T

Vectorization of LWR transient analysis code RELAP5/MOD1 and its effect

International Nuclear Information System (INIS)

Ishiguro, Misako; Harada, Hiroo; Shinozawa, Naohisa; Naraoka, Ken-itsu

1985-03-01

The RELAP5/MOD1 is a large thermal-hydraulic code to analyze LWR LOCA and non-LOCA transients. The code originally was designed for use on a CDC Cyber-176. This report documents vectorization of the RELAP5/MOD1 code conducted for the purpose of efficient use of VP-100 (peak speed 250 MFLOPS, clock period 7.5 ns) at the JAERI. The code was vectorized using the junction and volume level parallelisms in the hydrodynamic calculations, and the heat-structure and heat-mesh level in the heat conduction calculations. The vectorized version runs as much as 2.4 to 2.8 times faster than the original scalar version, while the speedup ratio is dependent on the number of spactial cells included in the problem. (author)

Modélisation de l'évaporation en milieu poreux: développement de modèles fondamentaux et appliqués

OpenAIRE

Debaste, Frédéric

2008-01-01

L'étude des phénomènes fondamentaux detransport et de thermodynamique apparaissant lors de l'évaporationen milieu poreux permet l'investigation d'applications pratiquesvariées. Dans ce travail, nous développons des modèles fondamentauxd'évaporation en milieu poreux que nous appliquons ensuite auséchage en lit fluidisé de deux matériaux granulaires poreux :lePVC et la levure.Les modèles mis au point sont réalisés suivant une approchemultiéchelle. Nous nous intéressons tout d'abord aux phénomèn...

Origin of the diversity in DNA recognition domains in phasevarion associated modA genes of pathogenic Neisseria and Haemophilus influenzae.

Science.gov (United States)

Gawthorne, Jayde A; Beatson, Scott A; Srikhanta, Yogitha N; Fox, Kate L; Jennings, Michael P

2012-01-01

Phase variable restriction-modification (R-M) systems have been identified in a range of pathogenic bacteria. In some it has been demonstrated that the random switching of the mod (DNA methyltransferase) gene mediates the coordinated expression of multiple genes and constitutes a phasevarion (phase variable regulon). ModA of Neisseria and Haemophilus influenzae contain a highly variable, DNA recognition domain (DRD) that defines the target sequence that is modified by methylation and is used to define modA alleles. 18 distinct modA alleles have been identified in H. influenzae and the pathogenic Neisseria. To determine the origin of DRD variability, the 18 modA DRDs were used to search the available databases for similar sequences. Significant matches were identified between several modA alleles and mod gene from distinct bacterial species, indicating one source of the DRD variability was via horizontal gene transfer. Comparison of DRD sequences revealed significant mosaicism, indicating exchange between the Neisseria and H. influenzae modA alleles. Regions of high inter- and intra-allele similarity indicate that some modA alleles had undergone recombination more frequently than others, generating further diversity. Furthermore, the DRD from some modA alleles, such as modA12, have been transferred en bloc to replace the DRD from different modA alleles.

Udvikling af vaccine mod dyr tarmsygdom

DEFF Research Database (Denmark)

Jungersen, Gregers

2012-01-01

DTU Veterinærinstituttet arbejder på at udvikle en effektiv vaccine mod bakterien Lawsonia intracellularis, der forårsager den antibiotikakrævende tarmsygdom proliferativ enteritis. Bakterien driller i laboratoriet, så forskerne må finde innovative veje til vaccinen. Målet er, at vaccinens...

Main pumps lost incident in the nuclear power plant Atucha I. Modelling with RELAP5/MOD3.2

International Nuclear Information System (INIS)

Ventura, M.A.; Rosso, R.D.

1998-01-01

Time evolution of natural circulation in the nuclear power plant Atucha I (CNA-I), in a main pumps lost incident because of the lost of external power feed, is analyzed. It leads to a strong stop transient, without an important blow down, from a forced nominal flow to a natural circulation one. The results are obtained from RELAP5/MOD3.2 code's modeling. The study is based on the refrigeration conditions analysis, during the first minutes of the reactor out of service. Previously to the transient, work had been done to obtain the plant steady state, with design parameters in operation conditions at 100 % of power. The object is that the actual plant state would be represented. In this way, each plant part (steam generators, reactor, pressurizer, pumps) had been modeled in separated form with the appropriate boundary conditions, to be used in the whole circuit simulation. The developed model, had been validated making use of the comparison between the values obtained to the principal thermodynamic parameters with the plant recorded values, in the same incident. The results are satisfactory in a way. On the other hand, it has suggested some modeling changes. The RELAP5/MOD3.2 capability to model the thermodynamic phenomena in a PHWR plant has been verified when, according to the mentioned incident, the flow pass from a nominal forced flow, to one which is governed by natural circulation, still with the CNA-I untypical design conditions. (author) [es

Obamas Fortsatte Krig mod Terror

DEFF Research Database (Denmark)

Ulrich, Philip Christian

2013-01-01

Kronikken argumenterer for at den type overvågningsskandaler som er fulgt i kølvandet på Edward Snowdens afsløringer blot er et symptom på den nye fase af krigen mod terror som Obama administrationen har ønsket at føre USA ind i. Den nye fase vil være præget af mere efterretningsvirksomhed snarere...

Beskyttelse mod passiv rygning i botilbud

DEFF Research Database (Denmark)

Faber, Louise

2018-01-01

ikke tidligere har været forbud. Denne problemstilling er aktuelt lige nu i forhold til passiv rygning, der tolereres mindre end tidligere. I denne artikel undersøges, i hvilket omfang der er mulighed for, at udstede rygeforbud på botilbud, når formålet med forbuddet er at beskytte andre beboere mod......Myndigheder har pligt til at beskytte individer mod ufrivillig sundhedsskadelig påvirkning i medfør af EMRK art. 8, stk. 1. På områder, hvor individets tolerancetærskel for sundhedsskadelig påvirkning bliver mindre end tidligere, kan der opstå behov for at udstede forbud i situationer, hvor der...... ufrivillig passiv rygning. Problemstilingen belyses i en forvaltningsretlig kontekst med udgangspunkt i dansk ret. Forfatteren konkluderer, at der er hjemmel i den danske Lov om Røgfrie Miljøer til at forbyde rygning, men at reglerne ikke giver tilstrækkelig beskyttelse i den situation, hvor en beboer...

ModSAF-based development of operational requirements for light armored vehicles

Science.gov (United States)

Rapanotti, John; Palmarini, Marc

2003-09-01

Light Armoured Vehicles (LAVs) are being developed to meet the modern requirements of rapid deployment and operations other than war. To achieve these requirements, passive armour is minimized and survivability depends more on sensors, computers, countermeasures and communications to detect and avoid threats. The performance, reliability, and ultimately the cost of these systems, will be determined by the technology trends and the rates at which they mature. Defining vehicle requirements will depend upon an accurate assessment of these trends over a longer term than was previously needed. Modelling and simulation are being developed to study these long-term trends and how they contribute to establishing vehicle requirements. ModSAF is being developed for research and development, in addition to the original requirement of Simulation and Modelling for Acquisition, Rehearsal, Requirements and Training (SMARRT), and is becoming useful as a means for transferring technology to other users, researchers and contractors. This procedure eliminates the need to construct ad hoc models and databases. The integration of various technologies into a Defensive Aids Suite (DAS) can be designed and analyzed by combining field trials and laboratory data with modelling and simulation. ModSAF (Modular Semi-Automated Forces,) is used to construct the virtual battlefield and, through scripted input files, a "fixed battle" approach is used to define and implement contributions from three different sources. These contributions include: models of technology and natural phenomena from scientists and engineers, tactics and doctrine from the military and detailed analyses from operations research. This approach ensures the modelling of processes known to be important regardless of the level of information available about the system. Survivability of DAS-equipped vehicles based on future and foreign technology can be investigated by ModSAF and assessed relative to a test vehicle. A vehicle can

PReMod: a database of genome-wide mammalian cis-regulatory module predictions.

Science.gov (United States)

Ferretti, Vincent; Poitras, Christian; Bergeron, Dominique; Coulombe, Benoit; Robert, François; Blanchette, Mathieu

2007-01-01

We describe PReMod, a new database of genome-wide cis-regulatory module (CRM) predictions for both the human and the mouse genomes. The prediction algorithm, described previously in Blanchette et al. (2006) Genome Res., 16, 656-668, exploits the fact that many known CRMs are made of clusters of phylogenetically conserved and repeated transcription factors (TF) binding sites. Contrary to other existing databases, PReMod is not restricted to modules located proximal to genes, but in fact mostly contains distal predicted CRMs (pCRMs). Through its web interface, PReMod allows users to (i) identify pCRMs around a gene of interest; (ii) identify pCRMs that have binding sites for a given TF (or a set of TFs) or (iii) download the entire dataset for local analyses. Queries can also be refined by filtering for specific chromosomal regions, for specific regions relative to genes or for the presence of CpG islands. The output includes information about the binding sites predicted within the selected pCRMs, and a graphical display of their distribution within the pCRMs. It also provides a visual depiction of the chromosomal context of the selected pCRMs in terms of neighboring pCRMs and genes, all of which are linked to the UCSC Genome Browser and the NCBI. PReMod: http://genomequebec.mcgill.ca/PReMod.

Culture and Creativity: World of Warcraft Modding in China and the US

Science.gov (United States)

Kow, Yong Ming; Nardi, Bonnie

Modding - end-user modification of commercial hardware and software - can be traced back at least to 1961 when Spacewar! was developed by a group of MIT students on a DEC PDP-1. Spacewar! evolved into arcade games including Space Wars produced in 1977 by Cinematronics (Sotamaa 2003). In 1992, players altering Wolfenstein 3-D (1992), a first person shooter game made by id Software, overwrote the graphics and sounds by editing the game files. Learning from this experience, id Software released Doom in 1993 with isolated media files and open source code for players to develop custom maps, images, sounds, and other utilities. Players were able to pass on their modifications to others. By 1996, with the release of Quake, end-user modifications had come to be known as "mods," and modding was an accepted part of the gaming community (Kucklich 2005; Postigo 2008a, b). Since late-2005, we have been studying World of Warcraft (WoW) in which the use of mods is an important aspect of player practice (Nardi and Harris 2006; Nardi et al. 2007). Technically minded players with an interest in extending the game write mods and make them available to players for free download on distribution sites. Most modders work for free, but the distribution sites are commercial enterprises with advertising.

Computation programs for the thermofluidodynamic transient analysis in the containment system following a LOCA

International Nuclear Information System (INIS)

Gorlandi, A.; Mazzini, M.; Oriolo, F.

1979-01-01

This works briefly describes the features of the computation codes available at the Istituto di Impianti Nucleari of the Pisa University for the analysis of the thermofluidodynamic transient in the containment system of a nuclear power plant following a LOCA (RELAP 4/MOD.S, COMPARE, FUMO and CONTEMPT-LT/026). More details are contained in the Annex. Particular attention has been devoted to the opportunity to study, through the computation codes, the effects of the sub division of a full pressure containment system

Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code; Implicacion de las capacidades de union fenosa dentro del area de termohidraulica en el APS de la C.N. Jose Cabrera. Aplicaciones del codigo RELAP5/MOD2

Energy Technology Data Exchange (ETDEWEB)

Martin, L; Saenz Tejada, P [Empresarios Agrupados, A.I.E., Madrid (Spain)

1993-12-15

When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

Flux pinning characteristics of Sn-doped YBCO film by the MOD process

International Nuclear Information System (INIS)

Choi, S.M.; Shin, G.M.; Yoo, S.I.

2013-01-01

Highlights: ► The pinning effects of undoped and Sn-doped YBCO films by MOD were characterized. ► Sn-containing nanoparticles were trapped in Sn-doped YBCO films by MOD. ► Sn-containing nanoparticles were identified as the YBa 2 SnO 5.5 (YBSO) phase by TEM. ► The YBSO nanoparticles are responsible for improved flux pinning effect. ► We report the orientation relationship between YBSO nanoparticles and YBCO matrix. -- Abstract: Compared with the undoped YBa 2 Cu 3 O 7−δ (YBCO) film, 10 mol% Sn-doped YBCO film exhibited significantly enhanced critical current densities (J c ) in magnetic fields up to 5 T at 65 and 77 K for H//c, indicating that the Sn-doped YBCO film possesses more effective flux pinning centers. Both samples were grown on the SrTiO 3 (STO) (1 0 0) single crystal substrates by the metal-organic deposition (MOD) process. Larger J c (77 K, 1 T) values of Sn-doped YBCO film are observed over a wide field-orientation angle (θ) except the field-orientations close to the ab-plane of YBCO (85° c values for 85° 2 SnO 5.5 (YBSO) phase by STEM (scanning transmission electron microscopy)-EDS (energy dispersive X-ray spectroscopy) analysis. Further analyses by HR-TEM (high resolution-transmission electron microscopy) revealed that YBSO nanoparticles completely surrounded by the YBCO matrix had random orientation with YBCO while those located at the interface of YBCO/STO substrate had epitaxial relationship with YBCO

Assessment of ICRF Antenna Performance in Alcator C-Mod

International Nuclear Information System (INIS)

Schilling, G.; Wukitch, S.J.; Lin, Y.; Basse, N.; Bonoli, P.T.; Edlund, E.; Lin, L.; Parisot, A.; Porkolab, M.

2004-01-01

The Alcator C-Mod has presented a challenge to install high-power ICRF antennas in a tight space. Modifications have been made to the antenna plasma-facing surfaces and the internal current-carrying structure in order to overcome performance limitations. At the present time, the antennas have exceeded 5 MW into plasma with heating phasing, up to 2.7 MW with current-drive phasing, with good efficiency and no deleterious effects

Modèle d’alerte des crises bancaires basé sur une approche hybride : modèle bayésien – machines à vecteurs supports

Directory of Open Access Journals (Sweden)

Taha Zaghdoudi

2016-09-01

Full Text Available Ces dernières années, la succession des crises bancaires, qui dans la plupart ont été soldées par des pertes économiques et financières énormes, a suscité l’intérêt de plusieurs chercheurs. Empiriquement, ces auteurs ont opté pour des modèles d’alerte précoce (Early Warning System pour prévenir leurs occurrences. L’objectif de ce papier est de construire un Modèle d’alerte des crises bancaires basé sur une approche hybride. Sur la base des données relatives à 22 pays qui ont subi des crises bancaires observées sur la période 1990–2011, nous avons développé un modèle d’alerte des crises bancaires. Ce modèle est basé sur une approche hybride Bayesian model averaging–Support vectors machine. Sur les 25 variables explicatives retenues, les résultats empiriques du modèle hybride ont fait ressortir 9 indicateurs qui sont considérés comme les principaux facteurs déterminants du déclenchement des crises bancaires. Ces derniers ont une probabilité postérieure d’inclusion supérieure à 0,5. Ces indicateurs potentiels sont la rentabilité nette des actifs, la compétitivité de l’intermédiation bancaire, les provisions sur les créances douteuses, les investissements directs étrangers, la concentration bancaire, la stabilité financière des banques, les produits nets financiers, le taux d’intérêt réel et le taux d’inflation.

Independent assessment of the TRAC-BD1/MOD1 computer code at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Wilson, G.E.; Charboneau, B.L.; Dallman, R.J.; Kullberg, C.M.; Wagner, K.C.; Wheatley, P.D.

1984-01-01

Under auspices of the United States Nuclear Regulatory Commission, their primary boiling water reactor safety analysis code (TRAC-BWR) is being assessed with simulations of a wide range of experimental data. The FY-1984 assessment activities were associated with the latest version (TRAC-BD1/MOD1) of this code. Typical results of the assessment studies are given. Conclusions formulated from these results are presented. These calculations relate to the overall applicability of the current code to safety analysis, and to future work which would further enhance the code's quality and ease of use

CONTEMPT 4/MOD 3: a multicompartment containment system analysis program

International Nuclear Information System (INIS)

Cheng, T.C.; Metcalfe, L.J.; Hartman, J.E.; Mings, W.J.; Crail, A.C.

1982-12-01

CONTEMPT4/MOD3 is a digital computer program, written in FORTRAN IV, that describes the behavior of multicompartment pressurized water reactor (PWR) containment systems and experimental containment systems subjected to postulated loss-of-coolant accident (LOCA) conditons. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and intercompartment mass and energy exchange based on user-supplied values for compartment descriptions, time step and edit controls, and selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, fan coolers, heat-conducting structures, sump drains, and PWR ice condensers. Dynamic stroage allocation (DSA) is used to limit the amount of computer core used for each problem. Optional automatic time step control allows the code to determine time step sizes within limits dictated by the user. Multicompartment capability (up to 999 individual compartments) and generalized, user-oriented input-data descriptions permit improved flexibility over previous codes in the CONTEMPT series. Analytical model descriptions, input instructions, and sample problem results are presented

Effect of reflection on Hα emissions in Alcator C-MOD

International Nuclear Information System (INIS)

Karney, C.F.; Stotler, D.P.; Skinner, C.H.; Terry, J.L.; Pappas, D.A.

1999-01-01

In order to explain anomalous intensity ratios which have been observed in Alcator C-MOD, the H α emissions in that experiment have been modeled with the DEGAS 2 code including the effects of wall reflection. By assuming that the first wall has different reflection coefficients for the two polarizations, we have qualitatively reproduced the observed anomaly. copyright 1999 American Institute of Physics

Preparation and properties of Y{sub 1-x}Ho{sub x}Ba{sub 2}Cu{sub 3}O{sub 7-{delta}} thin films by TFA-MOD method

Energy Technology Data Exchange (ETDEWEB)

Jian Hongbin [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Li Qi; Shi Dongqi [Institute for Superconducting and Electronic Materials, University of Wollongong, Wollongong 2522 (Australia); Zhang Li [Department of Mathematic and Physics, Anhui University of Architecture, Hefei 230022 (China); Yang Zhaorong [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Dou Shixue [Institute for Superconducting and Electronic Materials, University of Wollongong, Wollongong 2522 (Australia); Zhu Xuebin, E-mail: xbzhu@issp.ac.cn [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sun Yuping [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China)

2011-12-15

Y{sub 1-x}Ho{sub x}BCO thin films were prepared by TFA-MOD. The best performances were obtained for the Y{sub 0.6}Ho{sub 0.4}BCO thin film. The pinning mechanism was {delta}l-type for all derived thin films. Y{sub 1-x}Ho{sub x}Ba{sub 2}Cu{sub 3}O{sub 7-{delta}} (x = 0, 0.1, 0.2, 0.3, 0.4, 0.5) thin films were prepared on LaAlO{sub 3} (0 0 1) substrates by trifluoroacetate metal organic deposition (TFA-MOD) without change of the processing parameters. The highest J{sub c} was attributed to the sample of Y{sub 0.6}Ho{sub 0.4}Ba{sub 2}Cu{sub 3}O{sub 7-{delta}} thin film, whose critical current density is about 1.6 times as compared to that of YBa{sub 2}Cu{sub 3}O{sub 7-{delta}} thin film at 77 K and self field. The flux pinning type was not varied with Ho substitution and can be attributed to {delta}l pinning model, which is attributed to the close ionic radius between the Y{sup 3+} and Ho{sup 3+} ions. The improvement of J{sub c} by Ho substitution without change of the processing parameters will provide an effective route to enhance the J{sub c} of YBCO-based thin films using TFA-MOD method.

Dimensionering af stålrammebygninger mod kipning

DEFF Research Database (Denmark)

Borchersen, E.; Frederiksen, J.O.; Skov, K.

Rapporten beskriver en metode til dimensionering af stålrammebygninger mod kipning. Metoden er baseret dels på elasticitetsteoretiske overvejelser, dels på forsøg udført med 3-charniers stålrammmer i fuld skala....

Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code

International Nuclear Information System (INIS)

Martin, L.; Saenz Tejada, P.

1993-01-01

When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

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

Particle size distribution of dust collected from Alcator C-MOD

International Nuclear Information System (INIS)

Gorman, S.V.; Carmack, W.J.; Hembree, P.B.

1998-01-01

There are important safety issues associated with tokamak dust, accumulated primarily from sputtering and disruptions. The dust may contain tritium, it may be activated, chemically toxic, and chemically reactive. The purpose of this paper is to present results from analyses of particulate collected from the Alcator C-MOD tokamak located at Massachusetts Institute of Technology (MIT) in Cambridge, Massachusetts. The sample obtained from C-MOD was not originally intended for examination outside of MIT. The sample was collected with the intent of performing only a composition analysis. However, MIT provided the INEEL with this sample for particle analysis. The sample was collected by vacuuming a section of the machine (covering approximately 1/3 of the machine surface) with a coarse fiber filter as the collection surface. The sample was then analyzed using an optical microscope, SEM microscope, Microtrac FRA particle size analyzer. The data fit a log-normal distribution. The count median diameter (CMD) of the samples ranged from 0.3 microm to 1.1 microm with geometric standard deviations (GSD) ranging from 2.8 to 5.2 and a mass median diameter (MMD) ranging from 7.22 to 176 microm

Perspectives d’avenir du modèle autrichien

OpenAIRE

Neisser, Heinrich

2018-01-01

Plusieurs de ceux qui m’ont précédé ont déjà décrit des éléments de ce modèle autrichien qui suscite beaucoup d’intérêt à l’étranger, mais qui – et cela aussi a été signalé un certain nombre de fois – ne peut pas être transposé dans sa totalité à d’autres pays. De tels modèles concrets pour la solution de conflits sociaux naissent à partir d’un certain contexte historique. Ils se développent de manière pragmatique, c’est-à-dire pour résoudre, le mieux possible, des problèmes qui se posent à u...

RELAP5/MOD3 analysis of a heated channel in downflow

International Nuclear Information System (INIS)

Dimenna, R.A.; Qureshi, Z.H.; Boman, A.L.

1993-01-01

The onset of flow instability (OFI) is a significant phenomenon affecting the determination of a safe operating power limit in the Savannah River Site production reactors. Tests performed at Columbia University for a single tube with uniform axial and azimuthal heating have been analyzed with RELAP5/NPR, Version 0, a version of RELAP5/MOD3. The tests include water flow rates from 3.2 x l0 -4 - 2.l x 10 -3 m 3 /s (5 - 33 gpm), Reynolds numbers from 30,000 - 400,000, and surface heat fluxes from 0 - 3.2 x l0 6 w/m 2 (0 - 1,000,000 Btu/hr- ft 2 ). Pressure drop versus flow rate curves were mapped for both fixed pressure boundary conditions and fixed flow boundary conditions. RELAP5/MOD3 results showed fair agreement with data for both types of boundary conditions, and good internal consistency between calculations using the two different types of boundary conditions. Under single-phase unheated conditions, the code overpredicted the pressure drop by 22 - 34%. Under single-phase heated conditions, the overprediction increased to as much as 55%. For those tests where two-phase conditions were observed at the channel exit, RELAP5 predicted lower flows than seen in the tests before voiding occurred

Thermal-hydraulic analysis of the semiscale Mod-1 blowdown heat transfer test series

International Nuclear Information System (INIS)

Cozzuol, J.M.

1976-06-01

Selected experimental thermal-hydraulic data from the recent Semiscale Mod-1 blowdown heat transfer test series are analyzed from an experimental viewpoint with emphasis on explaining those phenomena which influence core fluid behavior. Comparisons are made between the trends measured by the system instrumentation and the trends predicted by the RELAP4 computer code to aid in obtaining an understanding of the interactions between phenomena occurring in different parts of the system. The analyses presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a pressurized water reactor during a postulated loss-of-coolant accident

3d Modeling of Combustion for Di-Si Engines Modélisation 3D de la combustion dans les moteurs à injection directe d'essence

Directory of Open Access Journals (Sweden)

Duclos J. P.

2006-12-01

Full Text Available Direct injection of gasoline is a promising concept to reduce fuel consumption of SI engines. The development of GDI engines is difficult and 3D CFD is a way to support its design. It requires models able to describe the spray and its evaporation and combustion. This paper presents a model, the ECFM, that enables to compute combustion for stratified load in the GDI engines. This model is a development of the Coherent Flame Model which includes thermal expansion effects, and is coupled with a burnt/unburnt gases conditionnal thermodynamic properties description. The model is validated by comparing measurements and computations on the GDI Mitsubishi engine in production. L'injection directe d'essence (IDE est un concept prometteur pour les moteurs à allumage commandé. La mise au point de ce type de moteur est néanmoins délicate, et le calcul 3D des chambres de combustion est un moyen d'aider à leur conception. Ceci nécessite cependant de disposer de modèles adaptés, à même de décrire le jet d'essence, son évaporation et la combustion du mélange créé. Cet article présente un modèle ECFM de simulation de la combustion dans les moteurs IDE, y compris en fonctionnement stratifié. C'est un développement du modèle flamme cohérente qui comprend des effets d'expansion thermique et est couplé avec une description conditionnelle gaz frais/gaz brûlés des grandeurs thermodynamiques. Ce modèle a été validé par rapprochement de mesures et simulations sur le moteur GDI Mitsubishi.

Assessment of a pressurizer spray valve faulty opening transient at Asco Nuclear Power Plant with RELAP5/MOD2

International Nuclear Information System (INIS)

Reventos, F.; Baptista, J.S.; Navas, A.P.; Moreno, P.

1993-12-01

The Asociacion Nuclear Asco has prepared a model of Asco NPP using RELAP5/MOD2. This model, which include thermalhydraulics, kinetics and protection and controls, has been qualified in previous calculations of several actual plant transients. One of the transients of the qualification process is a ''Pressurizer spray valve faulty opening'' presented in this report. It consists in a primary coolant depressurization that causes the reactor trip by overtemperature and later on the actuation of the safety injection. The results are in close agreement with plant data

Autres modèles de prestation de services publics dans les ...

International Development Research Centre (IDRC) Digital Library (Canada)

Les écrits sur la prestation de services publics à l'aide de modèles différents ont porté jusqu'à maintenant sur des endroits très précis. Ils ont par ailleurs été circonscrits à des secteurs donnés et ont manqué d'uniformité sur le plan méthodologique. Ce projet vise à analyser les modèles de prestation de services de santé, ...

Analysis of In-Vessel Late Phase Melt Progression Using SCDAP/RELAP5/MOD3.3

International Nuclear Information System (INIS)

Park, R.J.; Kim, S.B.; Kim, H.D.

2004-01-01

High-pressure in-vessel melt progressions of the KSNP (Korean Standard Nuclear Power Plant) have been analyzed using the SCDAP/RELAP5/MOD3.3 computer code. The total loss of feed water (LOFW) to the steam generators with/without intentional RCS depressurization using the safety depressurization system (SDS) and the station blackout (SBO) have been simulated from transient initiation to reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that the pressure boundary of the reactor coolant system did not fail before reactor vessel failure in the high-pressure sequences of the LOFW and the SBO transients of the KSNP. In all the high-pressure transients, approximately 20-30 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure. Intentional RCS depressurization using the SDS for the total LOFW delays reactor vessel failure for approximately 5 hours by actuation of the safety injection tanks. At the time of reactor vessel failure, approximately 50-60 % of the fuel rod cladding was oxidized for the total LOFW and the SBO transients of the KSNP. (authors)

Thermohydraulic analysis of nuclear power plant accidents by computer codes

International Nuclear Information System (INIS)

Petelin, S.; Stritar, A.; Istenic, R.; Gregoric, M.; Jerele, A.; Mavko, B.

1982-01-01

RELAP4/MOD6, BRUCH-D-06, CONTEMPT-LT-28, RELAP5/MOD1 and COBRA-4-1 codes were successful y implemented at the CYBER 172 computer in Ljubljana. Input models of NPP Krsko for the first three codes were prepared. Because of the high computer cost only one analysis of double ended guillotine break of the cold leg of NPP Krsko by RELAP4 code has been done. BRUCH code is easier and cheaper for use. Several analysis have been done. Sensitivity study was performed with CONTEMPT-LT-28 for double ended pump suction break. These codes are intended to be used as a basis for independent safety analyses. (author)

ModFOLD6: an accurate web server for the global and local quality estimation of 3D protein models.

Science.gov (United States)

Maghrabi, Ali H A; McGuffin, Liam J

2017-07-03

Methods that reliably estimate the likely similarity between the predicted and native structures of proteins have become essential for driving the acceptance and adoption of three-dimensional protein models by life scientists. ModFOLD6 is the latest version of our leading resource for Estimates of Model Accuracy (EMA), which uses a pioneering hybrid quasi-single model approach. The ModFOLD6 server integrates scores from three pure-single model methods and three quasi-single model methods using a neural network to estimate local quality scores. Additionally, the server provides three options for producing global score estimates, depending on the requirements of the user: (i) ModFOLD6_rank, which is optimized for ranking/selection, (ii) ModFOLD6_cor, which is optimized for correlations of predicted and observed scores and (iii) ModFOLD6 global for balanced performance. The ModFOLD6 methods rank among the top few for EMA, according to independent blind testing by the CASP12 assessors. The ModFOLD6 server is also continuously automatically evaluated as part of the CAMEO project, where significant performance gains have been observed compared to our previous server and other publicly available servers. The ModFOLD6 server is freely available at: http://www.reading.ac.uk/bioinf/ModFOLD/. © The Author(s) 2017. Published by Oxford University Press on behalf of Nucleic Acids Research.

A field evaluation of the Hardy TB MODS Kit™ for the rapid phenotypic diagnosis of tuberculosis and multi-drug resistant tuberculosis.

Directory of Open Access Journals (Sweden)

Laura Martin

Full Text Available Even though the WHO-endorsed, non-commercial MODS assay offers rapid, reliable TB liquid culture and phenotypic drug susceptibility testing (DST at lower cost than any other diagnostic, uptake has been patchy. In part this reflects misperceptions about in-house assay quality assurance, but user convenience of one-stop procurement is also important. A commercial MODS kit was developed by Hardy Diagnostics (Santa Maria, CA, USA with PATH (Seattle, WA, USA to facilitate procurement, simplify procedures through readymade media, and enhance safety with a sealing silicone plate lid. Here we report the results from a large-scale field evaluation of the MODS kit in a government service laboratory.2446 sputum samples were cultured in parallel in Lowenstein-Jensen (LJ, conventional MODS and in the MODS kit. MODS kit DST was compared with conventional MODS (direct DST and proportion method (indirect DST. 778 samples (31.8% were Mycobacterium tuberculosis culture-positive. Compared to conventional MODS the sensitivity, specificity, positive, and negative predictive values (95% confidence intervals of the MODS Kit were 99.3% (98.3-99.8%, 98.3% (97.5-98.8%, 95.8% (94.0-97.1%, and 99.7% (99.3-99.9%. Median (interquartile ranges time to culture-positivity (and rifampicin and isoniazid DST was 10 (9-13 days for conventional MODS and 8.5 (7-11 for MODS Kit (p<0.01. Direct rifampicin and isoniazid DST in MODS kit was almost universally concordant with conventional MODS (97.9% agreement, 665/679 evaluable samples and reference indirect DST (97.9% agreement, 687/702 evaluable samples.MODS kit delivers performance indistinguishable from conventional MODS and offers a convenient, affordable alternative with enhanced safety from the sealing silicone lid. The availability in the marketplace of this platform, which conforms to European standards (CE-marked, readily repurposed for second-line DST in the near future, provides a fresh opportunity for improving equity of

Quick look report for semiscale MOD-2C Test S-FS-11

International Nuclear Information System (INIS)

Plessinger, M.P.

1985-11-01

Results of a preliminary analysis of the fifth test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-11 simulated a pressurized water reactor transient initiated by a 50% break in a steam generator bottom feedwater line downstream of the check valve. With the exception of primary pressure, the initial conditions represented the initial conditions used for the C-E System 80 Final Safety Analysis Report (FSAR) Appendix 15B calculations. The transient included an initial 600 s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant followed by break isolation and affected loop steam generator refill with auxiliary feedwater. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overpressurization and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overpressurization and primary-to-secondary heat transfer. 64 figs

Control of substrate oxidation in MOD cerawwwmic coating on low-activation ferritic steel with reduced-pressure atmosphere

Science.gov (United States)

Tanaka, Teruya; Muroga, Takeo

2014-12-01

An Er2O3 ceramic coating fabricated using the metal-organic decomposition (MOD) method on a Cr2O3-covered low-activation ferritic steel JLF-1 substrate was examined to improve hydrogen permeation barrier performance of the coating. The Cr2O3 layer was obtained before coating by heat treating the substrate at 700 °C under reduced pressures of baking. Preprocessing to obtain a Cr2O3 layer would provide flexibility in the coating process for blanket components and ducts. Moreover, the Cr2O3 layer suppressed hydrogen permeation through the JLF-1 substrate. While further optimization of the coating fabrication process is required, it would be possible to suppress hydrogen permeation significantly by multilayers of Cr2O3 and MOD oxide ceramic.

Edge plasma physics modifications due to magnetic ripple in RFX-mod

International Nuclear Information System (INIS)

Scarin, P.; Agostini, M.; Carraro, L.; Cavazzana, R.; Ciaccio, G.; De Masi, G.; Spizzo, G.; Spolaore, M.; Vianello, N.

2015-01-01

The edge of the RFX-mod (R = 2 m, a = 0.46 m) Reversed Field Pinch is characterized by weak magnetic chaos affecting ion and electron diffusion. Edge particle transport is strongly influenced by a toroidal asymmetry caused by magnetic islands. An ambipolar radial electric field ensures local neutrality and possesses the same symmetry as the parent magnetic ripple: the result is the modulation of the perpendicular flow, with a slowing-down at the island X-point. In this paper we present a complete statistical analysis, over a large database of RFX-mod discharges, of the edge properties as they are modified by the magnetic topology: the plasma wall footprint follows the helical shape of the dominant central mode (m/n = 1/7), with an increase of H α emission and electron density corresponding to the O-point of the inner magnetic island. Edge turbulence is modified by the magnetic topology, being generated in the O-point region and damped near the X-point

The RANDOM computer program: A linear congruential random number generator

Science.gov (United States)

Miles, R. F., Jr.

1986-01-01

The RANDOM Computer Program is a FORTRAN program for generating random number sequences and testing linear congruential random number generators (LCGs). The linear congruential form of random number generator is discussed, and the selection of parameters of an LCG for a microcomputer described. This document describes the following: (1) The RANDOM Computer Program; (2) RANDOM.MOD, the computer code needed to implement an LCG in a FORTRAN program; and (3) The RANCYCLE and the ARITH Computer Programs that provide computational assistance in the selection of parameters for an LCG. The RANDOM, RANCYCLE, and ARITH Computer Programs are written in Microsoft FORTRAN for the IBM PC microcomputer and its compatibles. With only minor modifications, the RANDOM Computer Program and its LCG can be run on most micromputers or mainframe computers.

Betænkning om Indsatsen mod ungdomskriminalitet

DEFF Research Database (Denmark)

Reimann, Johan; Balvig, Flemming; Bay, Jens

Kommissionen foretager i betænkningen en samlet gennemgang af indsatsen mod ungdomskriminalitet og præsenterer på grundlag heraf en række forslag til, hvordan indsatsen kan styrkes med henblik på at gøre den så målrettet og virkningsfuld som mulig. Til brug for sine overvejelser har kommissionen...... overvejelser og forslag vedrørende indsatsen mod ungdomskriminalitet omfatter dels den forebyggende indsats, dels reaktionen på kriminalitet, der begås af børn og unge. På baggrund af erfaringerne fra eksisterende tiltag fremhæver kommissionen generelt den tidlige, helhedsorienterede, tværsektorielle og...... sammenhængende kriminalitetsforebyggende indsats som den væsentligste, hvis man for alvor ønsker at sætte effektivt ind for at begrænse ungdomskriminaliteten. Derudover stiller kommissionen en lang række forslag om konkrete tiltag med henblik på at styrke henholdsvis den forebyggende indsats og reaktionen på...

RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Foretic, D.

2003-01-01

In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

Combustion Modeling with the G-Equation Modélisation de la combustion avec l'équation de G

Directory of Open Access Journals (Sweden)

Peters N.

2006-12-01

Full Text Available Numerical investigations concerning the turbulent flame front propagation in Gasoline Direct Injection (GDI engines were made by implementing a flamelet model in the CFD code Fire. The advantage of this combustion model is the decoupling of the chemistry from the turbulent flow. For this purpose the combustion chamber has to be divided into a burned and an unburned area, which is realized by transporting a scalar field (G-Equation. The reference value defines the present averaged flame position. The complete reaction kinetics is calculated interactively with the CFD code in a one dimensional Representative Interactive Flamelet (RIF code. This combustion model was verified by simulating a 2. 0 l-2 V gasoline engine with homogeneous combustion where a parameter study was conducted to check the flamelet model for plausibility. Finally, the potential of this combustion model was investigated by simulating a hypothetical 2. 0 1-4 V GDI engine. Une investigation numérique relative à la propagation des fronts de flammes turbulents dans les moteurs à essence à injection directe (GDI a été menée en implantant un modèle de flameletdans le code 3D Fire. L'avantage de ce modèle de combustion est de découpler la chimie de l'écoulement turbulent en divisant la chambre de combustion en deux zones : brûlée et imbrûlée, à l'aide d'une équation de transport d'un scalaire (équation de G. Une valeur de référence de ce scalaire définit la position moyenne de la flamme. Une chimie complète est calculée interactivement avec le calcul 3D à l'aide d'un code monodimensionnel RIF (Representative Interactive Flamelet. Le modèle de combustion a été validé sur la simulation d'un moteur 2 litres à 2 soupapes en combustion homogène pour vérifier la représentativité de l'approche flamelet . Puis, le potentiel du modèle de combustion a été étudié en simulant un moteur modèle 2 litres 4 soupapes GDI.

RELAP5/MOD2 benchmarking study: Critical heat flux under low-flow conditions

International Nuclear Information System (INIS)

Ruggles, E.; Williams, P.T.

1990-01-01

Experimental studies by Mishima and Ishii performed at Argonne National Laboratory and subsequent experimental studies performed by Mishima and Nishihara have investigated the critical heat flux (CHF) for low-pressure low-mass flux situations where low-quality burnout may occur. These flow situations are relevant to long-term decay heat removal after a loss of forced flow. The transition from burnout at high quality to burnout at low quality causes very low burnout heat flux values. Mishima and Ishii postulated a model for the low-quality burnout based on flow regime transition from churn turbulent to annular flow. This model was validated by both flow visualization and burnout measurements. Griffith et al. also studied CHF in low mass flux, low-pressure situations and correlated data for upflows, counter-current flows, and downflows with the local fluid conditions. A RELAP5/MOD2 CHF benchmarking study was carried out investigating the performance of the code for low-flow conditions. Data from the experimental study by Mishima and Ishii were the basis for the benchmark comparisons

Plasma performance and scaling laws in the RFX-mod reversed-field pinch experiment

International Nuclear Information System (INIS)

Innocente, P.; Alfier, A.; Canton, A.; Pasqualotto, R.

2009-01-01

The large range of plasma currents (I p = 0.2-1.6 MA) and feedback-controlled magnetic boundary conditions of the RFX-mod experiment make it well suited to performing scaling studies. The assessment of such scaling, in particular those on temperature and energy confinement, is crucial both for improving the operating reversed-field pinch (RFP) devices and for validating the RFP configuration as a candidate for the future fusion reactors. For such a purpose scaling laws for magnetic fluctuations, temperature and energy confinement have been evaluated in stationary operation. RFX-mod scaling laws have been compared with those obtained from other RFP devices and numerical simulations. The role of the magnetic boundary has been analysed, comparing discharges performed with different active control schemes of the edge radial magnetic field.

Investigation of an inventory calculation model for a solvent extraction system and the development of its computer programme - SEPHIS-J

International Nuclear Information System (INIS)

Ihara, Hitoshi; Nishimura, Hideo; Ikawa, Koji; Ido, Masaru.

1986-11-01

In order to improve the applicability of near-real-time materials accountancy (N.R.T.MA) to a reprocessing plant, it is necessary to develop an estimation method for the nuclear material inventory at a solvent extraction system under operation. For designing the solvent extraction system, such computer codes as SEPHIS, SOLVEX and TRANSIENTS had been used. Accuracy of these codes in tracing operations and predicting inventories in the extraction system had been discussed. Then, much better codes, e.g., SEPHIS Mod4 and PUBG, were developed. Unfortunately, SEPHIS Mod4 was not available in countries other than the USA and PUBG was not suitable for use with a mini-computer which would be practical as a field computer because of quite a lot of computing time needed. The authors investigated an inventory estimation model compatible with PUBG in functions and developed the corresponding computer programme, SEPHIS-J, based on the SEPHIS Mod3 code, resulting in a third of computing time compared with PUBG. They also validated the programme by calculating a static state as well as a dynamic one of the solvent extraction process and by comparing them among the programme, SEPHIS Mod3 and PUBG. Using the programme, it was shown that the inventory changes due to changes of feed flow and concentration were not so small that they might be neglected although the changes of feed flow and concentration were within measurement errors. (author)

MOD-AGE - an algorithm for age-depth model construction; U-series dated speleothems case study

Science.gov (United States)

Hercman, H.; Pawlak, J.

2012-04-01

We present MOD-AGE - a new system for chronology construction. MOD-AGE can be used for profiles that have been dated by different methods. As input data, the system uses the following basic measurements: activities, atomic ratios or age, as well as depth measurement. Based on probability distributions describing the measurement results, MOD-AGE estimates the age~depth relation and its confidence bands. To avoid the use of difficult-to-meet assumptions, MOD-AGE uses nonparametric methods. We applied a Monte Carlo simulation to model age and depth values based on the real distribution of counted data (activities, atomic ratios, depths etc.). Several fitting methods could be applied for estimating the relationships; based on several tests, we decide to use LOESS method (locally weighted scatterplot smoothing). The stratigraphic correction procedure applied in the MOD-AGE program uses a probability calculus, which assumes that the ages of all the samples are correctly estimated. Information about the probability distribution of the samples' ages is used to estimate the most probable sequence that is concordant according to the superposition rule. MOD-AGE is presented as a tool for the chronology construction of speleothems that have been analyzed by the U-series method, and it is compared to the StalAge algorithm presented by D. Scholtz and D.L Hoffmann (2011). Scholtz, D., Hoffmann, D. L., 2011. StalAge - An algorithm designed for construction of speleothem age models. Quaternary Geochronology 6, 369-382.

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.

Edge Turbulence Imaging on NSTX and Alcator C-Mod

International Nuclear Information System (INIS)

S.J. Zweben; R.A. Maqueda; J.L. Terry; B. Bai; C.J. Boswell; C.E. Bush; D. D'Ippolito; E.D. Fredrickson; M. Greenwald; K. Hallatschek; S. Kaye; B. LaBombard; R. Maingi; J. Myra; W.M. Nevins; B.N. Rogers; D.P. Stotler; J. Wilgen; and X.Q. Xu

2002-01-01

Edge turbulence images have been made using an ultra-high speed CCD camera on both NSTX and Alcator C-Mod. In both cases, the D-alpha or HeI (587.6 nm) line emission from localized deuterium or helium gas puffs was viewed along a local magnetic field line near the outer midplane. Fluctuations in this line emission reflect fluctuations in electron density and/or electron temperature through the atomic excitation rates, which can be modeled using the DEGAS-2 code. The 2-D structure of the measured turbulence can be compared with theoretical simulations based on 3-D fluid models

Calculation of pre and post-test of the third. proposed standard problem exercise, for the PMK-NVH-IAEA experiment using the RELAP4/MOD5 and RELAP5/MOD1

International Nuclear Information System (INIS)

Neves Conti, T. das; Sabundjian, G.; Oliveira Neto, J.M. de

1992-01-01

The results of RELAP4/MOD5 and RELAP5/MOD1 modeling tests against the steam generator tube rupture experiments performed at PMK-NVH Experimental Loop Facility (IAEA-Standard Problem Exercise-3) are presented in the report. The pre and post-test results, when compared against the experimental data were satisfactorily good, except a discrepancy in the steam-generator relief valve opening time. (author)

Uncertainty in RELAP5/MOD3.2 calculations for interfacial drag in downward two-phase flow

International Nuclear Information System (INIS)

Clark, Collin; Schlegel, Joshua P.; Hibiki, Takashi; Ishii, Mamoru; Kinoshita, Ikuo

2016-01-01

Highlights: • Uncertainty propagation is key for best estimate code reliability. • Uncertainty in drift flux correlations used to evaluate uncertainty in interfacial drag. • Bias and error have been compared for various models. - Abstract: RELAP5/MOD3.2 is a thermal-hydraulic system analysis code used to predict the response of nuclear reactor coolant systems in the event of certain accident scenarios. It is important that RELAP and other system analysis codes are able to accurately predict various two-phase flow phenomena, particularly the interfacial transfers between the liquid and gas phases. It is also important to understand how much uncertainty exists in these predictions due to uncertainties in the constitutive relations used to close the two-fluid model. In this paper, the uncertainty in the interfacial drag calculated by RELAP5/MOD3.2 due to errors in the drift-flux models used to close the model is evaluated and compared to the correlation developed by Goda et al. (2003). The case of downward flow is considered due to the importance of co-current and counter-current downward flow for predicting behavior in the downcomer of reactor systems during small-break Loss of Coolant Accidents (LOCAs) in nuclear reactor systems. The overall uncertainty in the interfacial force calculations due to error in the distribution parameter models were found to have a bias of +8.1% and error of 20.1% for the models used in RELAP5, and a bias of −30.8% and error of 23.1% for the correlation of Goda et al. (2003). However this analysis neglects the effects of compensating errors in the drift-flux parameters, as the drift velocity is assumed to be perfectly accurate. More physically meaningful results could be obtained if the distribution parameter and drift velocity were calculated directly from local phase concentration and velocity measurements, however no studies were available which included all of this information.

Three-dimensional Simulation of Gas Conductance Measurement Experiments on Alcator C-Mod

International Nuclear Information System (INIS)

Stotler, D.P.; LaBombard, B.

2004-01-01

Three-dimensional Monte Carlo neutral transport simulations of gas flow through the Alcator C-Mod subdivertor yield conductances comparable to those found in dedicated experiments. All are significantly smaller than the conductance found with the previously used axisymmetric geometry. A benchmarking exercise of the code against known conductance values for gas flow through a simple pipe provides a physical basis for interpreting the comparison of the three-dimensional and experimental C-Mod conductances

Detailed Post Analysis of HERMES-HALF Experiment using RELAP5/MOD3

Energy Technology Data Exchange (ETDEWEB)

Park, Rae Joon; Kang, Kyung Ho; Ha, Kwang Soon; Cho, Young Ro; Koo, Kil Mo; Kim, Sang Baik; Kim, Hee Dong

2005-03-15

As part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of APR1400, a HERMES-HALF experiment has been analyzed to verify and evaluate the experimental results using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is very similar to the experimental results of the HERMES-HALF, in general. Increases in the water inlet area and the water level in the reactor cavity lead to an increase in the water circulation mass flow rate. In the small water inlet area condition, the lower value of the water outlet area has an effect on the water circulation mass flow rate, but the larger value of this has no effect. The air injection mass flow rate has no effect on the water circulation mass flow rate when it is greater than 40 % at the small water inlet area condition. However, an increase in the air injection mass flow rate leads to an increase in the water circulation mass flow rate. In the large water inlet area condition, increases in the water outlet area and the air injection mass flow rate lead to an increase in the water circulation mass flow rate. As the water outlet moves to a lower position, the water circulation mass flow rate slowly increases.

Assessment study of RELAP5/MOD2, CYCLE 36. 04 based on spray start-up test for DOEL-4

Energy Technology Data Exchange (ETDEWEB)

Moeyaert, P.; Stubbe, E.

1989-07-01

This report presents an assessment study for the code RELAP-5 MOD-2 based on a pressurizer spray start-up test of the Doel-4 power plant. Doel-4 is a three loop WESTINGHOUSE PWR plant ordered by the EBES utility with a nominal power rating of 1000 MWe and equipped with preheater type E steam generators. A large series of commissioning tests are normally performed on new plants, of which the so called pressurizer spray and heater test (SU-PR-01) was performed on February 2nd 1985. TRACTEBEL, being the Architect-Engineer for this plant was closely involved with all start-up tests and was responsible for the final approval of the tests.

Disruption Neutral Point Experiment on Alcator C-Mod

Science.gov (United States)

Granetz, R. S.; Nakamura, Y.

2000-10-01

Disruptions of single-null elongated plasmas generally result in loss of vertical position control, leading to a current quench occurring at the top or bottom of the machine, with all the attendant problems of halo and eddy currents flowing in divertor structures. On JT-60U, it has been found that if the plasma is operated with its magnetic axis at a particular height, called the neutral point, the initial vertical drift after a thermal quench is significantly slower than usual, and sometimes can even be arrested, thereby avoiding a current quench in the divertor region entirely. In an ongoing collaboration between MIT and JAERI, the neutral point concept is being tested in Alcator C-Mod, which has a significantly higher plasma elongation than JT-60U (1.65 vs 1.3). Calculations using TSC predict a neutral point at z~=+1 cm above the midplane (a=22 cm). The existence of a neutral point has now been experimentally confirmed, albeit at a height of z=+2.7 cm. The plasma has remained vertically stable for up to 9 ms after the disruption thermal quench, which in principle, is long enough for the PF control system to respond, if programmed appropriately. In addition, the physics of the neutral point stability on C-Mod appears to be somewhat different than that on JT-60U.

Results of Semiscale Mod-2C small-break (5%) loss-of-coolant accident. Experiments S-LH-1 and S-LH-2

International Nuclear Information System (INIS)

Loomis, G.G.; Streit, J.E.

1985-11-01

Two experiments simulating small break (5%) loss-of-coolant accidents (5% SBLOCAs) were performed in the Semiscale Mod-2C facility. These experiments were identical except for downcomer-to-upper-head bypass flow (0.9% in Experiment S-LH-1 and 3.0% in Experiment S-LH-2) and were performed at high pressure and temperature [15.6 MPa (2262 psia) system pressure; 37 K (67 0 F) core differential temperature; 595 K(610 0 F) hot leg fluid temperature]. From the experimental results, the signature response and transient mass distribution are determined for a 5% SBLOCA. The core thermal-hydraulic response is characterized, including core void distribution maps, and the effect of core bypass flow on transient severity is assessed. Comparisons are made between postexperiment RELAP5 calculations and the experimental results, and the capability of RELAP5 to calculate the phenomena is assessed. 115 figs

Parallel transport studies of high-Z impurities in the core of Alcator C-Mod plasmas

Energy Technology Data Exchange (ETDEWEB)

Reinke, M. L.; Hutchinson, I. H.; Rice, J. E.; Greenwald, M.; Howard, N. T.; Hubbard, A.; Hughes, J. W.; Terry, J. L.; Wolfe, S. M. [MIT-Plasma Science and Fusion Center Cambridge, Massachusetts 02139 (United States)

2013-05-15

Measurements of poloidal variation, ñ{sub z}/, in high-Z impurity density have been made using photodiode arrays sensitive to vacuum ultraviolet and soft x-ray emission in Alcator C-Mod plasmas. In/out asymmetries in the range of −0.2<0.3 are observed for r/a<0.8, and accumulation on both the high-field side, n{sub z,cos}<0, and low-field side, n{sub z,cos}>0, of a flux surface is found to be well described by a combination of centrifugal, poloidal electric field, and ion-impurity friction effects. Up/down asymmetries, −0.05<0.10, are observed over 0.50 corresponding to accumulation opposite the ion ∇B drift direction. Measurements of the up/down asymmetry of molybdenum are found to disagree with predictions from recent neoclassical theory in the trace limit, n{sub z}Z{sup 2}/n{sub i}≪1. Non-trace levels of impurities are expected to modify the main-ion poloidal flow and thus change friction-driven impurity density asymmetries and impurity poloidal rotation, v{sub θ,z}. Artificially modifying main-ion flow in parallel transport simulations is shown to impact both ñ{sub z}/ and v{sub θ,z}, but simultaneous agreement between measured and predicted up/down and in/out asymmetry as well as impurity poloidal rotation is not possible for these C-Mod data. This link between poloidal flow and poloidal impurity density variation outlines a more stringent test for parallel neoclassical transport theory than has previously been performed. Measurement and computational techniques specific to the study of poloidal impurity asymmetry physics are discussed as well.

CONTAIN LMR/1B-Mod.1, A computer code for containment analysis of accidents in liquid-metal-cooled nuclear reactors

International Nuclear Information System (INIS)

Murata, K.K.; Carroll, D.E.; Bergeron, K.D.; Valdez, G.D.

1993-01-01

The CONTAIN computer code is a best-estimate, integrated analysis tool for predicting the physical, chemical, and radiological conditions inside a nuclear reactor containment building following the release of core material from the primary system. CONTAIN is supported primarily by the U. S. Nuclear Regulatory Commission (USNRC), and the official code versions produced with this support are intended primarily for the analysis of light water reactors (LWR). The present manual describes CONTAIN LMR/1B-Mod. 1, a code version designed for the analysis of reactors with liquid metal coolant. It is a variant of the official CONTAIN 1.11 LWR code version. Some of the features of CONTAIN-LMR for treating the behavior of liquid metal coolant are in fact present in the LWR code versions but are discussed here rather than in the User's Manual for the LWR versions. These features include models for sodium pool and spray fires. In addition to these models, new or substantially improved models have been installed in CONTAIN-LMR. The latter include models for treating two condensables (sodium and water) simultaneously, sodium atmosphere and pool chemistry, sodium condensation on aerosols, heat transfer from core-debris beds and to sodium pools, and sodium-concrete interactions. A detailed description of each of the above models is given, along with the code input requirements

Analysis of the OECD-LOFT International Standard Problem 31 using SCDAP/RELAP5/MOD3

International Nuclear Information System (INIS)

Hohorst, J.K.; Allison, C.M.

1992-01-01

The CORA-13 bundle heating and melting experiment performed at the Kernforechungszentrum, Karlaruhe, (KfK) was analyzed at the Idaho National Engineering Laboratory (INEL) using SCDAP/RELAP5/MOD3. This analysis was part of a systematic assessment of SCDAP/RELAP5/MOD3 for the US Nuclear Regulatory Commission to (a) evaluate the variances between calculated and observed behavior, (b) identify outstanding modeling deficiencies, and (c) to evaluate the impact of ongoing modeling improvements. A brief discussion of the CORA-13 experiment including a description of the facility, important test conditions, and comparisons with other CORA experimental conditions and results is provided in this report. This report describes the results of the SCDAP/RELAPS/MOD3 analysis including a description of the SCDAP/RELAPS model of the facility, base case results, sensitivity results, and a comparison with other SCDAP/RELAP5/MOD3 code-to-data comparisons

Simulation de l'habitat physique du barbeau fluviatile (Barbus barbus, L. 1758 : choix des modèles biologiques et sensibilité de la réponse

Directory of Open Access Journals (Sweden)

POUILLY M.

1994-07-01

Full Text Available Des courbes monovariées et des modèles multivariés de préférence d'habitat du barbeau, Barbus barbus, ont été établis à partir de données récoltées sur 3 cours d'eau. Dans les deux cas (monovarié et multivarié, trois modèles locaux, correspondant aux données d'une rivière, et un modèle général, regroupant l'ensemble des données, ont été établis. La qualité de la prédiction et la sensibilité de la réponse lors de la simulation des capacités d'accueil d'un cours d'eau révèlent : 1 que les modèles multivariés ont une valeur prédictive plus forte que les courbes de préférence, 2 que les modèles construits à partir de jeux de données locaux sont plus performants que les modèles généraux, et 3 que la perte de précision est moindre dans le cas du modèle général multivarié.

Implementation of PWR steady state self-initialization feature into RELAP4/MOD6/U4/J3

International Nuclear Information System (INIS)

Yoshida, Kazuo

1987-07-01

A PWR steady state self-initialization feature has been implemented into the RELAP4/MOD6/U4/J3 code which is an improved version of RELAP4/MOD6 and can analyze not only large break but also small break LOCA in LWRs. This feature is originated from RELAP4/MOD7 which is the most updated released version of RELAP4 from INEL. Several FORTRAN subroutines in MOD7 related to this feature were transplanted into MOD6/U4/J3 with some improvements, which were the modification of method to take a balance of heat transfer between primary and secondary side at SG-U tubes, and to make it possible to nodalize secondary side of SG as multi-node. Advantages realized by implementation of this option are saving of time in initializaing a new model and an assurance of steady state and self consistency of input data in a small break LOCA analysis of a PWR. (author)

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)

Developmental assessment of RELAP5/MOD3 code against ROSA-IV/TPTF horizontal two-phase flow experiments

International Nuclear Information System (INIS)

Kukita, Yutaka; Asaka, Hideaki; Anoda, Yoshinari; Ishiguro, Misako; Tasaka, Kanji; Mimura, Yuichi; Nemoto, Toshiyuki.

1990-03-01

A developmental version of the RELAP5/Mod3 code (as of June 1989) was assessed for accuracy using experimental data taken for high-pressure (7MPa) steam-water two-phase flow in a large-diameter (0.18 m) horizontal-pipe test section of the ROSA-IV Two-Phase Flow Test Facility (TPTF). The agreement between the measured and calculated test section void fractions was much better than that for the previous generation of RELAP5 (MOD2). The improvement was achieved primarily due to the code changes with respect to the flow stratification criterion and interfacial-drag calculation scheme. (author)

Analyzing the loss of coolant accident in PWR nuclear reactors with elevation change in cold leg by RELAP5/MOD3.2 system code

International Nuclear Information System (INIS)

Kheshtpaz, H.; Alison, C.

2006-01-01

As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the RELAP5/MOD3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect

Rotation and transport in Alcator C-Mod ITB plasmas

Science.gov (United States)

Fiore, C. L.; Rice, J. E.; Podpaly, Y.; Bespamyatnov, I. O.; Rowan, W. L.; Hughes, J. W.; Reinke, M.

2010-06-01

Internal transport barriers (ITBs) are seen under a number of conditions in Alcator C-Mod plasmas. Most typically, radio frequency power in the ion cyclotron range of frequencies (ICRFs) is injected with the second harmonic of the resonant frequency for minority hydrogen ions positioned off-axis at r/a > 0.5 to initiate the ITBs. They can also arise spontaneously in ohmic H-mode plasmas. These ITBs typically persist tens of energy confinement times until the plasma terminates in radiative collapse or a disruption occurs. All C-Mod core barriers exhibit strongly peaked density and pressure profiles, static or peaking temperature profiles, peaking impurity density profiles and thermal transport coefficients that approach neoclassical values in the core. The strongly co-current intrinsic central plasma rotation that is observed following the H-mode transition has a profile that is peaked in the centre of the plasma and decreases towards the edge if the ICRF power deposition is in the plasma centre. When the ICRF resonance is placed off-axis, the rotation develops a well in the core region. The central rotation continues to decrease as long as the central density peaks when an ITB develops. This rotation profile is flat in the centre (0 ITB density profile is observed (0.5 ITB foot that is sufficiently large to stabilize ion temperature gradient instabilities that dominate transport in C-Mod high density plasmas.

BEACON/MOD2A analysis of the Arkansas-1 reactor cavity during a hypothetical hot leg break

International Nuclear Information System (INIS)

Ramsthaler, J.A.

1979-01-01

As part of the evaluation of the new MOD2A version of the BEACON code, the Arkansas-1 reactor cavity was modeled during a hypothetical loss-of-coolant accident. Results of the BEACON analysis were compared with results obtained previously with the COMPARE containment code. Studies were also made investigating some of the BEACON interphasic, timestep control, and wall heat transfer options to assure that these models were working properly and to observe their effects on the results. Descriptions of the Arkansas-1 reactor cavity, initial assumptions during the hypothetical LOCA, and methods of modeling with BEACON are presented. Some of the problems encountered in accurately modeling the penetrations surrounding the hot and cold leg pipes are also discussed

Calculation of design load for the MOD-5A 7.3 mW wind turbine system

Science.gov (United States)

Mirandy, L.; Strain, J. C.

1995-01-01

Design loads are presented for the General Electric MOD-SA wind turbine. The MOD-SA system consists of a 400 ft. diameter, upwind, two-bladed, teetered rotor connected to a 7.3 mW variable-speed generator. Fatigue loads are specified in the form of histograms for the 30 year life of the machine, while limit (or maximum) loads have been derived from transient dynamic analysis at critical operating conditions. Loads prediction was accomplished using state of the art aeroelastic analyses developed at General Electric. Features of the primary predictive tool - the Transient Rotor Analysis Code (TRAC) are described in the paper. Key to the load predictions are the following wind models: (1) yearly mean wind distribution; (2) mean wind variations during operation; (3) number of start/shutdown cycles; (4) spatially large gusts; and (5) spatially small gusts (local turbulence). The methods used to develop statistical distributions from load calculations represent an extension of procedures used in past wind programs and are believed to be a significant contribution to Wind Turbine Generator analysis. Test/theory correlations are presented to demonstrate code load predictive capability and to support the wind models used in the analysis. In addition MOD-5A loads are compared with those of existing machines. The MOD-5A design was performed by the General Electric Company, Advanced Energy Program Department, under Contract DEN3-153 with NASA Lewis Research Center and sponsored by the Department of Energy.

64 Application d'un modèle conceptuel et d'un modèle de réseaux ...

African Journals Online (AJOL)

PR BOKO

Application of a conceptual model and a model of artificial neural networks for the ... évaluer l'impact hydrologique d'un aménagement d'un bassin ou pour ... une catégorie de modèles pluie–débit basée sur l'intelligence artificielle [4]. ..... et D.P. SOLOMATINE, « River flow forecasting using Artificial Neuronal Networks”.

Identification of functional elements and regulatory circuits by Drosophila modENCODE.

Science.gov (United States)

Roy, Sushmita; Ernst, Jason; Kharchenko, Peter V; Kheradpour, Pouya; Negre, Nicolas; Eaton, Matthew L; Landolin, Jane M; Bristow, Christopher A; Ma, Lijia; Lin, Michael F; Washietl, Stefan; Arshinoff, Bradley I; Ay, Ferhat; Meyer, Patrick E; Robine, Nicolas; Washington, Nicole L; Di Stefano, Luisa; Berezikov, Eugene; Brown, Christopher D; Candeias, Rogerio; Carlson, Joseph W; Carr, Adrian; Jungreis, Irwin; Marbach, Daniel; Sealfon, Rachel; Tolstorukov, Michael Y; Will, Sebastian; Alekseyenko, Artyom A; Artieri, Carlo; Booth, Benjamin W; Brooks, Angela N; Dai, Qi; Davis, Carrie A; Duff, Michael O; Feng, Xin; Gorchakov, Andrey A; Gu, Tingting; Henikoff, Jorja G; Kapranov, Philipp; Li, Renhua; MacAlpine, Heather K; Malone, John; Minoda, Aki; Nordman, Jared; Okamura, Katsutomo; Perry, Marc; Powell, Sara K; Riddle, Nicole C; Sakai, Akiko; Samsonova, Anastasia; Sandler, Jeremy E; Schwartz, Yuri B; Sher, Noa; Spokony, Rebecca; Sturgill, David; van Baren, Marijke; Wan, Kenneth H; Yang, Li; Yu, Charles; Feingold, Elise; Good, Peter; Guyer, Mark; Lowdon, Rebecca; Ahmad, Kami; Andrews, Justen; Berger, Bonnie; Brenner, Steven E; Brent, Michael R; Cherbas, Lucy; Elgin, Sarah C R; Gingeras, Thomas R; Grossman, Robert; Hoskins, Roger A; Kaufman, Thomas C; Kent, William; Kuroda, Mitzi I; Orr-Weaver, Terry; Perrimon, Norbert; Pirrotta, Vincenzo; Posakony, James W; Ren, Bing; Russell, Steven; Cherbas, Peter; Graveley, Brenton R; Lewis, Suzanna; Micklem, Gos; Oliver, Brian; Park, Peter J; Celniker, Susan E; Henikoff, Steven; Karpen, Gary H; Lai, Eric C; MacAlpine, David M; Stein, Lincoln D; White, Kevin P; Kellis, Manolis

2010-12-24

To gain insight into how genomic information is translated into cellular and developmental programs, the Drosophila model organism Encyclopedia of DNA Elements (modENCODE) project is comprehensively mapping transcripts, histone modifications, chromosomal proteins, transcription factors, replication proteins and intermediates, and nucleosome properties across a developmental time course and in multiple cell lines. We have generated more than 700 data sets and discovered protein-coding, noncoding, RNA regulatory, replication, and chromatin elements, more than tripling the annotated portion of the Drosophila genome. Correlated activity patterns of these elements reveal a functional regulatory network, which predicts putative new functions for genes, reveals stage- and tissue-specific regulators, and enables gene-expression prediction. Our results provide a foundation for directed experimental and computational studies in Drosophila and related species and also a model for systematic data integration toward comprehensive genomic and functional annotation.

MOD control center automated information systems security evolution

Science.gov (United States)

Owen, Rich

1991-01-01

The role of the technology infusion process in future Control Center Automated Information Systems (AIS) is highlighted. The following subject areas are presented in the form of the viewgraphs: goals, background, threat, MOD's AISS program, TQM, SDLC integration, payback, future challenges, and bottom line.

Vold mod førskolebørn

DEFF Research Database (Denmark)

Oldrup, Helene; Lindstrøm, Maia; Korzen, Sara

Denne rapport handler om praksis og barrierer for opsporing af og underretning om vold mod førskolebørn. Rapporten er baseret på 22 kvalitative interviews med fagfolk inden for dagpasning, sundhedsvæsen og det sociale system. Undersøgelsen viser bl.a., at en del af fagfolkene ofte tøver med at un...

RETRAN02/MOD02: an outside perspective

International Nuclear Information System (INIS)

Wei, T.Y.C.

1984-03-01

ANL recently participated in a review of the RETRAN02/MOD02 code to determine the range of accuracy, the reliability and the reproducibility of results obtained with the code for Chapter 15 non-LOCA system transients for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). This paper summarizes the technical aspects of that review

Evaluation of MODS Culture in the Diagnosis of Pulmonary Tuberculosis

Directory of Open Access Journals (Sweden)

Z Aminzadeh

2012-05-01

Full Text Available

Background and Objectives

Culture of M. tuberculosis is the golden standard for the diagnosis of TB which is a much more sensitive test than Smear examination. There is a strong need to use the new assays in order to speed up diagnostic methods. The aim of this research was to determine the evaluation of Microscopic Observation Drug Susceptibility culture in pulmonary tuberculosis in comparison with Ziehl-Neelsen stain and Lowenstein-Jensen culture of sputum.

Methods

The research method was a Cross-sectional (diagnostic test and the technique was observational-interview type. If the patient's history revealed clinical criteria compatible with TB and the infectious specialist’s judgment was that of "TB suspected case, the patient was considered a pulmonary TB suspect. Then, in addition to sputum Ziehl-Neelsen stain and culture for Lowenstein-Jensen, we carried out MODS culture as well.

Results

100 patients (48 male, 52 female with mean age of 52.9 ± 21.83 were evaluated. During sputum examination, 40% were Ziehl-Neelsen stain positive while 30% had positive sputum culture for Mycobacterium Tuberculosis in Lowenstein-Jensen and 47% had positive MODS culture. In comparison with sputum smear and Lowenstein-Jensen culture, MODS had a sensitivity of 82.5% and 86%, specificity of 77% and 70%, positive predictive value of 70% and 55%, negative predictive value of 86% and 92%, respectively.

Conclusion

Chromate Binding and Removal by the Molybdate-Binding Protein ModA.

Science.gov (United States)

Karpus, Jason; Bosscher, Michael; Ajiboye, Ifedayo; Zhang, Liang; He, Chuan

2017-04-04

Effective and cheap methods and techniques for the safe removal of hexavalent chromate from the environment are in increasingly high demand. High concentrations of hexavalent chromate have been shown to have numerous harmful effects on human biology. We show that the E. coli molybdate-binding protein ModA is a genetically encoded tool capable of removing chromate from aqueous solutions. Although previously reported to not bind chromate, we show that ModA binds chromate tightly and is capable of removing chromate to levels well below current US federal standards. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Modélisation multi-échelles en viscoplasticité endommageable de composites thermoplastiques renforcés par des fibres discontinues

OpenAIRE

ACHOUR, Nadia; CHATZIGEORGIOU, George; BONNAY, Kevin; MERAGHNI, Fodil

2017-01-01

Un nouveau modèle multi-échelles en régime viscoplastique endommageable est développé pour un composite à matrice polypropylène renforcé par des fibres de verre courtes. Basé sur l’approche en champs moyens de Mori Tanaka, il intègre une matrice viscoplastique modélisée par un modèle phénoménologique nommé par ses auteurs DSGZ et des fibres de verres modélisées par un comportement élastique linéaire. Le modèle multi-échelles permet d’intégrer la microstructure du composite préalablement carac...

Élaborer un modèle d'équilibre général dynamique et stochastique ...

International Development Research Centre (IDRC) Digital Library (Canada)

Un éventail d'outils employés en économie appliquée, comme les modèles de série chronologique ou les modèles économétriques qui s'appuient sur des propriétés ... Ces dernières années, des recherches ont montré que les modèles estimés d'EGDS peuvent produire des prévisions d'une plus grande exactitude que les ...

A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

International Nuclear Information System (INIS)

Kim, Sung Hyun; Moon, Chan Ki; Park, Sung Baek; Na, Man Gyun

2013-01-01

The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study

Validation of neutral point on JT-60U, Alcator C-Mod and ASDEX-Upgrade tokamaks

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Yoshino, Ryuji; Pautasso, Gabriella; Gruber, Otto; Jardin, Stephen

2002-01-01

Validation studies of a neutrally balanced vertical plasma position, so-called ''neutral point'', have been carried out by computational simulations and experiments under trilateral Japan-US-EU collaborations. It was clarified that the neutral point, where VDEs (Vertical Displacement Events) are hardly occurred, does exit in the Alcator C-Mod and ASDEX-Upgrade tokamaks as well as the JT-60U, consistent with the simulations. Meanwhile, precise details of the VDE behavior exhibit their own characters according to the individual of the tokamaks such as an up-down asymmetry of plasma shape. Sensitivity of the neutral point to the plasma shape and current profile was also addressed in detail. (author)

RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation

Directory of Open Access Journals (Sweden)

Andrej Prošek

2018-01-01

Full Text Available The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in light water reactors. The plant measured data come mostly from startup tests and operational events. Also, for operational events the measured plant data may not be sufficient to explain all details of the event. The purpose of this study was therefore besides code assessment to demonstrate that simulations can be very beneficial for deep understanding of the plant response and further corrective measures. The abnormal event with reactor trip and safety injection signal actuation was simulated with the latest RELAP5/MOD3.3 Patch 05 best-estimate thermal hydraulic computer code. The measured and simulated data agree well considering the major plant system responses and operator actions. This suggests that the RELAP5 code simulation is good representative of the plant response and can complement not available information from plant measured data. In such a way, an event can be better understood.

Transport Studies in Alcator C-Mod ITB Plasmas

Science.gov (United States)

Fiore, C. L.; Bonoli, P. T.; Ernst, D.; Greenwald, M. J.; Ince-Cushman, A.; Lin, L.; Marmar, E. S.; Porkolab, M.; Rice, J. E.; Wukitch, S.; Rowan, W.; Bespamyatnov, I.; Phillips, P.

2008-11-01

Internal transport barriers occur in C-Mod plasmas that have off-axis ICRF heating and also in Ohmic H-mode plasmas. These ITBs are marked by highly peaked density and pressure profiles, as they rely on a reduction of particle and thermal flux in the barrier region which allows the neoclassical pinch to peak the central density without reducing the central temperature. Enhancement of several core diagnostics has resulted in increased understanding of C-Mod ITBs. Ion temperature profile measurements have been obtained using an innovative design for x-ray crystal spectrometry and clearly show a barrier forming in the ion temperature profile. The phase contrast imaging (PCI) provides limited localization of the ITB related fluctuations that increase in strength as the central density increases. Simulation of triggering conditions, integrated simulations with fluctuation measurements, parametric studies, and transport implications of fully ionized boron impurity profiles in the plasma are under study. A summary of these results will be presented.

Small break LOCA RELAP5/MOD3 uncertainty quantification: Bias and uncertainty evaluation for important phenomena

International Nuclear Information System (INIS)

Ortiz, M.G.; Ghan, L.S.; Vogl, J.

1991-01-01

The Nuclear Regulatory Commission (NRC) revised the Emergency Core Cooling System (ECCS) licensing rule to allow the use of Best Estimate (BE) computer codes, provided the uncertainty of the calculations are quantified and used in the licensing and regulation process. The NRC developed a generic methodology called Code Scaling, Applicability and Uncertainty (CSAU) to evaluate BE code uncertainties. The CSAU methodology was demonstrated with a specific application to a pressurized water reactor (PWR), experiencing a postulated large break loss-of-coolant accident (LBLOCA). The current work is part of an effort to adapt and demonstrate the CSAU methodology to a small break (SB) LOCA in a PWR of B and W design using RELAP5/MOD3 as the simulation tool. The subject of this paper is the Assessment and Ranging of Parameters (Element 2 of the CSAU methodology), which determines the contribution to uncertainty of specific models in the code

Modèles d'affaires ouverts : nouveaux mécanismes de revenus pour ...

International Development Research Centre (IDRC) Digital Library (Canada)

Ce projet de recherche vise à examiner les principaux modèles d'affaires qui émergent dans une société réseautée, les répercussions qu'ont ces modèles sur les droits des consommateurs et des créateurs et le rôle qu'ils peuvent jouer pour favoriser une plus grande inclusion et la diversité culturelle. Le projet se penchera ...

Standardissimo. Les limitations théoriques du Modèle Standard. Quelles réponses y apporter?

Science.gov (United States)

Renard, F. M.

Nous présentons I 'état du Modèle Standard des interactions fortes, faibles et électromagnétiques. Après une description rapide de ses 3 secteurs, secteur de jauge (radiation), secteur fermionique (matière) et secteur scalaire (génération des masses), nous insistons sur le grand nombre de paramètres libres et sur les choix arbitraires qu'il a fallu faire dans l'élaboration du modèle. Nous faisons ressortir les problèmes techniques non résolus et nous dressons la liste des questions fondamentales restées sans réponses. Nous passons ensuite en revue les idées et méthodes proposées pour répondre à ces questions. Elles utilisent essentiellement 3 voies différentes. La première consiste à requérir plus de symétrie (extension du modèle, symétrie Gauche-Droite, Grandes Unifications, Supersymétrie,...). La seconde contient les diverses alternatives au Modèle Standard impliquant des modifications dans certains secteurs (par exemple le secteur scalaire avec le modèle de la Technicouleur) ou de façon plus violente l'hypothèse d'une sous-structure des leptons, des quarks et des bosons W et Z eux-mêmes. Une dernière voie cherche à justifier les particularités du Modèle Standard et relier ses paramètres libres en se basant sur des principes de cohérence interne du modèle. Les conséquences observables de ces diverses approches sont dans chaque cas mentionnées.

Kinetic modeling of divertor heat load fluxes in the Alcator C-Mod and DIII-D tokamaks

Energy Technology Data Exchange (ETDEWEB)

Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

2015-09-15

The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

Mise en place d'une base de données pour une modélisation ...

African Journals Online (AJOL)

Mise en place d'une base de données pour une modélisation hydrologique distribuée du bassin versant du Bandama (Côte d'Ivoire) : apport d'un modèle numérique d'altitude, de la télédétection et du SIG Physitel.

Assessment of RELAP5/MOD3.2.2γ against flooding database in horizontal-to-inclined pipes

International Nuclear Information System (INIS)

Kim, Hyoung Tae; No, Hee Cheon

2001-01-01

A total of 356 experimental data for the onset of flooding are compiled for the data bank and used for the assessment of RELAP5/MOD3.2.2γ predictions of Counter-Current Flow Limitation (CCFL) in horizontal-to-inclined pipes simulating a PWR hot leg. The predictions of the flooding gas velocity in the database are known to be largely dependent on the horizontal pipe length-to-diameter ratio (L/D). RELAP5 calculations are compared with the experimental data where L/D is varied within the range of database. The present input model used for the simulation of CCFL is validated to reasonably calculate the gradient of water level in the horizontal pipes connected with the inclined volumes. RELAP5 calculations show that the RELAP5 predicts the flooding points qualitatively well but higher gas flow rate is required to initiate the flooding compared with the experimental data if the L/D is as low that of the hot legs of typical PWRs. Standard RELAP5 code is modified to apply the user specified CCFL curve not only to veritical volumes but also to the horizontal volumes. The calculation value by the modified version lies well on the applied CCFL curve even if flooding occurs at lower gas velocity thatn predicted by the CCFL curve in standard RELAP5

A Unified Model for Slug Flow Generation Modélisation de la formation des bouchons : vers un modèle stochastique unifié

Directory of Open Access Journals (Sweden)

Bernicot M.

2006-11-01

essentiellement les lois de formation, l'évolution des bouchons faisant encore l'objet de travaux théoriques et de campagnes de tests effectués à l'Institut de Mécanique des Fluides de Toulouse, dans le cadre du Projet EVE associant Institut Français du Pétrole (IFP, Elf et Total. Notations utilisées : Les modèles mathématiques proposés sont basés sur: - Une identification des bouchons à partir de leur moment et de leur lieu d'apparition; pour le bouchon : Tn, Xn. - Le suivi dans le temps de la position du talon du n-ième bouchon à l'instant t: Yn (t et de sa longueur : Delta indice (n (t. - L'hypothèse d'un temps de coalescence Sn très important; tous les bouchons formés continuent d'exister (Delta indice (n (t > 0 au temps t : Tn < t < Sn; on exclut ici toute possibilité de coalescence dans la zone de formation des bouchons. - L'utilisation de processus de Poisson tenant compte de diverses hypothèses simplificatrices permettant d'approcher les différents mécanismes de formation de bouchons : -zones autorisées et interdites pour la formation des bouchons successifs; -vitesses de déplacement du talon des bouchons (avec ou sans phase d'accélération initiale; -influence de l'importance du glissement entre phases; -influence de la position de la zone de formation de bouchons par rapport à l'entrée de la conduite. . . 1 Modèle Bernicot-Drouffe Partant du modèle Bernicot-Drouffe présenté initialement à l'OTC de Houston en 1988 [4], nous commençons par en établir une justification mathématique rigoureuse après en avoir rappelé les hypothèses de base. Ces hypothèses qui sont strictement valables pour des écoulements en conduites horizontales et à proximité de l'entrée de la conduite (fig. 2, supposent essentiellement : - Un faible glissement entre phases et donc des vitesses de bouchons assimilables à des constantes pendant la phase de formation (la phase d'accélération initiale étant ignorée : (*** - Aucune formation de bouchon devant le

Improving containment mass and energy releases for CONTEMPT-LT/028 TU with RELAP5/MOD3

International Nuclear Information System (INIS)

DaSilva, H.C.; Choe, W.G.

1996-01-01

In order to obtain boundary conditions for RELAP5/MOD3 best estimate (BE) large break (LB) loss-of-coolant accident (LOCA) calculations, it is necessary to utilize a separate containment analysis code CONTEMPT-LT/028 TU, which in turn accepts mass and energy releases from the RELAP5/MOD3 calculation. When these boundary conditions are obtained, they are observed to be significantly lower than those reported in FSAR containment analyses. This motivates the present study, where RELAP5/MOD3 mass and energy releases are generated using the same assumptions listed in the FSAR containment calculations. Then CONTEMPT-LT/028 TU pressures and temperatures calculated with both sets of mass and energy releases are compared. It is seen that those obtained with the RELAP5/MOD3 input are still significantly lower, indicating a level of conservatism in the FSAR mass and energy releases that is even above that explicitly listed and also incorporated into the RELAP5/MOD3 calculation. An important conclusion from this finding is that Environmental Qualification (EQ) issues requiring containment re-analyses are likely to be easily resolved if new mass and energy releases are calculated with state-of-the-art LOCA codes modeling the entire reactor coolant system, even when conservative assumptions are incorporated

How sex- and age-disaggregated data and gender and generational analyses can improve humanitarian response.

Science.gov (United States)

Mazurana, Dyan; Benelli, Prisca; Walker, Peter

2013-07-01

Humanitarian aid remains largely driven by anecdote rather than by evidence. The contemporary humanitarian system has significant weaknesses with regard to data collection, analysis, and action at all stages of response to crises involving armed conflict or natural disaster. This paper argues that humanitarian actors can best determine and respond to vulnerabilities and needs if they use sex- and age-disaggregated data (SADD) and gender and generational analyses to help shape their assessments of crises-affected populations. Through case studies, the paper shows how gaps in information on sex and age limit the effectiveness of humanitarian response in all phases of a crisis. The case studies serve to show how proper collection, use, and analysis of SADD enable operational agencies to deliver assistance more effectively and efficiently. The evidence suggests that the employment of SADD and gender and generational analyses assists in saving lives and livelihoods in a crisis. © 2013 The Author(s). Journal compilation © Overseas Development Institute, 2013.

Assessment and improvement of condensation model in RELAP5/MOD3

Energy Technology Data Exchange (ETDEWEB)

Rho, Hui Cheon; Choi, Kee Yong; Park, Hyeon Sik; Kim, Sang Jae [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Lee, Sang Il [Korea Power Engineering Co., Inc., Seoul (Korea, Republic of)

1997-07-15

The objective of this research is to remove the uncertainty of the condensation model through the assessment and improvement of the various heat transfer correlations used in the RELAP5/MOD3 code. The condensation model of the standard RELAP5/MOD3 code is systematically arranged and analyzed. A condensation heat transfer database is constructed from the previous experimental data on various condensation phenomena. Based on the constructed database, the condensation models in the code are assessed and improved. An experiment on the reflux condensation in a tube of steam generator in the presence of noncondensable gases is planned to acquire the experimental data.

Correlation ECE diagnostic in Alcator C-Mod

International Nuclear Information System (INIS)

Sung, C.; Irby, J.; Leccacorvi, R.; Vieira, R.; Oi, C.; Rice, J.; Reinke, M.; Gao, C.; Ennever, P.; Porkolab, M.; Churchill, R.; Theiler, C.; Walk, J.; Hughes, J.; Hubbard, A.; Greenwald, M.

2015-01-01

Correlation ECE (CECE) is a diagnostic technique that allows measurement of small amplitude electron temperature, Te, fluctuations through standard cross-correlation analysis methods. In Alcator C-Mod, a new CECE diagnostic has been installed[Sung RSI 2012], and interesting phenomena have been observed in various plasma conditions. We find that local Te fluctuations near the edge (ρ ~ 0:8) decrease across the linearto- saturated ohmic confinement transition, with fluctuations decreasing with increasing plasma density [Sung NF 2013], which occurs simultaneously with rotation reversals [Rice NF 2011]. Te fluctuations are also reduced across core rotation reversals with an increase of plasma density in RF heated L-mode plasmas, which implies that the same physics related to the reduction of Te fluctuations may be applied to both ohmic and RF heated L-mode plasmas. In I-mode plasmas, we observe the reduction of core Te fluctuations, which indicates changes of turbulence occur not only in the pedestal region but also in the core across the L/I transition [White NF 2014]. The present CECE diagnostic system in C-Mod and these experimental results are described in this paper

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

International Nuclear Information System (INIS)

Szczurek, J.

1995-01-01

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

Energy Technology Data Exchange (ETDEWEB)

Szczurek, J. [Inst. of Atomic Energy, Swierk (Poland)

1995-12-31

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

Energy Technology Data Exchange (ETDEWEB)

Szczurek, J [Inst. of Atomic Energy, Swierk (Poland)

1996-12-31

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

Review of the SCDAP/RELAP5/MOD3.1 code structure and core T/H model before core damage

International Nuclear Information System (INIS)

Kim, See Darl; Kim, Dong Ha

1998-04-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code is being developed at the INEL under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. NRC. As The current time, the SCDAP/RELAP5/MOD3.1 code is the result of merging the RELAP5/MOD3 and SCDAP models. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. Major purpose of the report is to provide information about the characteristics of SCDAP/RELAP5/MOD3.1 core T/H models for an integrated severe accident computer code being developed under the mid/long-term project. This report analyzes the overall code structure which consists of the input processor, transient controller, and plot file handler. The basic governing equations to simulate the thermohydraulics of the primary system are also described. As the focus is currently concentrated in the core, core nodalization parameters of the intact geometry and the phenomenological subroutines for the damaged core are summarized for the future usage. In addition, the numerical approach for the heat conduction model is investigated along with heat convection model. These studies could provide a foundation for input preparation and model improvement. (author). 6 refs., 3 tabs., 4 figs

Transport and Stability in C-Mod ITBs in Diverse Regimes

Science.gov (United States)

Fiore, C. L.; Ernst, D. R.; Howard, N. T.; Kasten, C. P.; Mikkelsen, D.; Reinke, M. L.; Rice, J. E.; White, A. E.; Rowan, W. L.; Bespamyatnov, I.

2012-10-01

Internal Transport Barriers (ITBs) in C-Mod feature highly peaked density and pressure profiles and are typically induced by the introduction of radio frequency power in the ion cyclotron range of frequencies (ICRF) with the second harmonic of the resonance for minority hydrogen ions positioned off-axis at the plasma half radius on either the low or high field side of the plasma. These ITBs are formed in the absence of particle or momentum injection, and with monotonic q profiles with qminITB dynamics in a reactor relevant regime. Recently, linear and non-linear gyrokinetic simulations have demonstrated that changes in the ion temperature and plasma rotation profiles, coincident with the application of off-axis ICRF heating, contribute to greater stability to ion temperature gradient driven fluctuation in the plasma. This results in reduced turbulent driven outgoing heat flux. To date, ITB formation in C-Mod has only been observed in EDA H-mode plasmas with moderate (2-3 MW) ICRF power. Experiments to explore the formation of ITBs in other operating regimes such as I-mode and also with high ICRF power are being undertaken to understand further the process of ITB formation and sustainment, especially with regard to turbulent driven transport.

Validation of the RALOC-mod.4 thermal-hydraulics code on evaporation transients in the Phebus containment

International Nuclear Information System (INIS)

Spitz, P.B.; Lemoine, F.; Tirini, S.

1997-01-01

IPSN (Nuclear Protection and Safety Institute) and GRS (Gesellschaft fur Anlagen und Reaktorsicherheit Schwertnergasse 1) are developing the ESCADRE-ASTEC systems of codes devoted to the prediction of the behaviour of water-cooled reactors during a severe accident. The RALOC-mod 4 code belongs to this system and is specifically devoted to containment thermal-hydraulics studies. IPSN has designed a Thermal Hydraulic Containment Test Program in support to the Phebus Fission Product Test Program/2/. Evaporation tests have been recently performed in the Phebus containment test facility. The objective of this work is to assess against these tests the capability of the RALOC -mod 4 code to capture the phenomena observed in these experiments and more particularly the evaporation heat transfer and wall heat transfers. (DM)

The measurement of the intrinsic impurities of molybdenum and carbon in the Alcator C-Mod tokamak plasma using low resolution spectroscopy

Science.gov (United States)

May, M. J.; Finkenthal, M.; Regan, S. P.; Moos, H. W.; Terry, J. L.; Goetz, J. A.; Graf, M. A.; Rice, J. E.; Marmar, E. S.; Fournier, K. B.; Goldstein, W. H.

1997-06-01

The intrinsic impurity content of molybdenum and carbon was measured in the Alcator C-Mod tokamak using low resolution, multilayer mirror (MLM) spectroscopy ( Delta lambda ~1-10 AA). Molybdenum was the dominant high-Z impurity and originated from the molybdenum armour tiles covering all of the plasma facing surfaces (including the inner column, the poloidal divertor plates and the ion cyclotron resonant frequency (ICRF) limiter) at Alcator C-Mod. Despite the all metal first wall, a carbon concentration of 1 to 2% existed in the plasma and was the major low-Z impurity in Alcator C-Mod. Thus, the behaviour of intrinsic molybdenum and carbon penetrating into the main plasma and the effect on the plasma must be measured and characterized during various modes of Alcator C-Mod operation. To this end, soft X-ray extreme ultraviolet (XUV) emission lines of charge states, ranging from hydrogen-like to helium-like lines of carbon (radius/minor radius, r/a~1) at the plasma edge to potassium to chlorine-like (0.4Data Nucl. Data Tables 33 (1985) 149), which were incorporated into the collisional radiative model. The intrinsic i

Simulation of LOFT anticipated-transient experiments L6-1, L6-2, and L6-3 using TRAC-PF1/MOD1

International Nuclear Information System (INIS)

Sahota, M.S.

1984-01-01

Anticipated-transient experiments L6-1, L6-2, and L6-3, performed at the Loss-of-fluid Test (LOFT) facility, are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1/MOD1). The results are used to assess TRAC-PF1/MOD1 trip and control capabilities, and predictions of thermal-hydraulic phenomena during slow transients. Test L6-1 simulated a loss-of-stream load in a large pressurized-water reactor (PWR), and was initiated by closing the main steam-flow control valve (MSFCV) at its maximum rate, which reduced the heat removal from the secondary-coolant system and increased the primary-coolant system pressure that initiated a reactor scram. Test L6-2 simulated a loss-of-primary coolant flow in a large PWR, and was initiated by tripping the power to the primary-coolant pumps (PCPs) allowing the pumps to coast down. The reduced primary-coolant flow caused a reactor scram. Test L6-3 simulated an excessive-load increase incident in a large PWR, and was initiated by opening the MSFCV at its maximum rate, which increased the heat removal from the secondary-coolant system and decreased the primary-coolant system pressure that initiated a reactor scram. The TRAC calculations accurately predict most test events. The test data and the calculated results for most parameters of interest also agree well

Experiment data report for Semiscale Mod-1 Test S-05-1 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-1 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-05-1 was conducted from initial conditions of 2263 psia and 544 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 vessel lower plenum to simulate emergency core coolant injection in a PWR, with the flow rate based on system volume scaling

Power oscillation of the Mod-0 wind turbine

Science.gov (United States)

Seidel, R. C.

1978-01-01

The Mod-0 power has noise components with varying frequency patterns. Magnitudes reach more than forty percent power at the frequency of twice per rotor revolution. Analysis of a simple torsional model of the power train predicts less than half the observed magnitude and does not explain the shifting frequencies of the noise patterns.

Modélisation du comportement et des couplages HMC des milieux poreux

OpenAIRE

Hoang , Ha

2012-01-01

Modelling of the behavior and the couplings HMC of the porous circles; La modélisation du comportement hydromécanique chimique des milieux poreux saturés et non saturés est abordée au niveau microscopique et mésoscopique. Au niveau microscopique la modélisation des écoulements diphasiques est basée sur une représentation du réseau poral comme un ensemble de tubes dont les orientations et les rayons sont choisis sur un principe d’équivalence avec les pores. L’algorithme régissant la génération...

Modélisation hydro-thermique 2D d'un produit fortement déformable lors du séchage convectif

OpenAIRE

Hassini , Lamine; Azzouz , Soufiene; Belghith , Ali

2007-01-01

International audience; Le but de ce travail est de simuler en 2D l'évolution de la teneur en eau, de la température, de la taille et de la forme géométrique d'un produit fortement déformable lors du séchage convectif. Le modèle écrit dans un repère fixe, consiste en une équation de conservation de la phase solide, une équation de diffusion/convection de l'eau liquide et une équation de conduction/convection de chaleur, couplées par la vitesse de contraction de la phase solide due au retrait....

Investigation of RF-enhanced plasma potentials on Alcator C-Mod

International Nuclear Information System (INIS)

Ochoukov, R.; Whyte, D.G.; Brunner, D.; Cziegler, I.; LaBombard, B.; Lipschultz, B.; Myra, J.; Terry, J.; Wukitch, S.

2013-01-01

Radio frequency (RF) sheath rectification is a leading mechanism suspected of causing anomalously high erosion of plasma facing materials in RF-heated plasmas on Alcator C-Mod. An extensive experimental survey of the plasma potential (Φ P ) in RF-heated discharges on C-Mod reveals that significant Φ P enhancement (>100 V) is found on outboard limiter surfaces, both mapped and not mapped to active RF antennas. Surfaces that magnetically map to active RF antennas show Φ P enhancement that is, in part, consistent with the recently proposed slow wave rectification mechanism. Surfaces that do not map to active RF antennas also experience significant Φ P enhancement, which strongly correlates with the local fast wave intensity. In this case, fast wave rectification is a leading candidate mechanism responsible for the observed enhancement

High performance discharges and capabilities in Alcator C-Mod

International Nuclear Information System (INIS)

Porkolab, M.

1996-01-01

Alcator C-Mod is a compact, diverted, shaped, high magnetic field (B = 9 T) tokamak operating at the Massachusetts Institute of Technology Plasma Fusion Center. The machine interior is all metallic, and the walls and divertor region are covered with molybdenum tiles. The vacuum vessel is a continuous, thick wall stainless steel construction, prototypical of future fusion devices (e.g., ITER). Typical discharge cleaning utilizes ECDC, or electron-cyclotron discharge cleaning, in the steady state at low magnetic field (0.0875 T). While its dimensions are compact (R = 0.67 m, a = 0.22 m, K = 1.8), C-Mod is designed to operate up to 2.5 MA at 9.0 T magnetic field. To present date the machine has operated at currents up to 1.5 MA at B = 5.3 T, and magnetic fields up to 8.0 T at I p = 1.2 MA. Due to the high current density, line average densities of 4.0 x 10 20 m -3 are obtained with gas fueling, and peak densities in excess of 1.0 x 10 21 m -3 have been obtained with pellet fueling. Typical pulse lengths are up to 2.0 seconds, with a flat-top of typically 1.0 sec. Presently the device is equipped with 4.0 MW of ICRF heating power operating at 80 MHz, but this capability is being upgraded to 8.0 MW with the addition of 4.0 MW of tunable ICRF power operating at 40.80 MHz. A 20 pellet/pulse deuterium injector is operational, and a 4 pellet Li injector is also operational. To reduce the influx of metallic impurities during high power operation, recently boronization of the machine interior was begun prior to plasma discharges, this allowed plasma operation with full auxiliary power capability without excessive radiative power losses from the plasma core. 7 refs

Study of MOD control system in ECRH

International Nuclear Information System (INIS)

Su Yu; Liu Baohua; Ding Tonghai; Kuang Guangli

2005-01-01

High-voltage power supply (HVPS) is one of the important components in ECRH (Electron Cyclotron Resonance Heating). The MOD (modulator) control system is a key of the operation of HVPS and the whole system. The background and principium is introduced in this paper, especially the detail of the hardware and software of the control system is shown. The experiment, that shows stability, accuracy and reliability had reached the expected goal. (authors)

Comparison of Measurements and FluorMOD Simulations for Solar Induced Chlorophyll Fluorescence and Reflectance of a Corn Crop under Nitrogen Treatments [SIF and Reflectance for Corn

Science.gov (United States)

Middleton, Elizabeth M.; Corp, Lawrence A.; Campbell, Petya K. E.

2007-01-01

The FLuorescence Explorer (FLEX) satellite concept is one of six semifinalist mission proposals selected in 2006 for pre-Phase studies by the European Space Agency (ESA). The FLEX concept proposes to measure passive solar induced chlorophyll fluorescence (SIF) of terrestrial ecosystems. A new spectral vegetation Fluorescence Model (FluorMOD) was developed to include the effects of steady state SIF on canopy reflectance. We used our laboratory and field measurements previously acquired from foliage and canopies of corn (Zea mays L.) under controlled nitrogen (N) fertilization to parameterize and evaluate FluorMOD. Our data included biophysical properties, fluorescence (F) and reflectance spectra for leaves; reflectance spectra of canopies and soil; solar irradiance; plot-level leaf area index; and canopy SIF emissions determined using the Fraunhofer Line Depth principal for the atmospheric telluric oxygen absorption features at 688 nm (O2-beta) and 760 nm (O2-alpha). FluorMOD simulations implemented in the default "look-up-table" mode did not reproduce the observed magnitudes of leaf F, canopy SIF, or canopy reflectance. However, simulations for all of these parameters agreed with observations when the default FluorMOD information was replaced with measurements, although N treatment responses were underestimated. Recommendations were provided to enhance FluorMOD's potential utility in support of SIF field experiments and studies of agriculture and ecosystems.

Experiments and Simulations of ITER-like Plasmas in Alcator C-Mod

International Nuclear Information System (INIS)

Wilson, R.; Kessel, C.E.; Wolfe, S.; Hutchinson, I.H.; Bonoli, P.; Fiore, C.; Hubbard, A.E.; Hughes, J.; Lin, Y.; Ma, Y.; Mikkelsen, D.; Reinke, M.; Scott, S.; Sips, A.C.C.; Wukitch, S.

2010-01-01

Alcator C-Mod is performing ITER-like experiments to benchmark and verify projections to 15 MA ELMy H-mode Inductive ITER discharges. The main focus has been on the transient ramp phases. The plasma current in C-Mod is 1.3 MA and toroidal field is 5.4 T. Both Ohmic and ion cyclotron (ICRF) heated discharges are examined. Plasma current rampup experiments have demonstrated that (ICRF and LH) heating in the rise phase can save voltseconds (V-s), as was predicted for ITER by simulations, but showed that the ICRF had no effect on the current profile versus Ohmic discharges. Rampdown experiments show an overcurrent in the Ohmic coil (OH) at the H to L transition, which can be mitigated by remaining in H-mode into the rampdown. Experiments have shown that when the EDA H-mode is preserved well into the rampdown phase, the density and temperature pedestal heights decrease during the plasma current rampdown. Simulations of the full C-Mod discharges have been done with the Tokamak Simulation Code (TSC) and the Coppi-Tang energy transport model is used with modified settings to provide the best fit to the experimental electron temperature profile. Other transport models have been examined also.

Participer à l’évolution du modèle québécois

Directory of Open Access Journals (Sweden)

Manuel Dussault

2002-10-01

Full Text Available Ce court texte vise à expliquer la position de l’Association des manufacturiers et exportateurs du Québec (AMEQ sur la question du modèle québécois. Pour l’association, il existe un modèle québécois, comme il existe un modèle français, américain, japonais ou autres. Ce modèle, ni tout a fait original, ni tout à fait pareil aux autres, est à la fois libéral et social démocratique. Il se caractérise par le nationalisme, l’équité et la concertation. L’histoire économique moderne est devenue, en partie celle du rôle de l’État par les leviers économiques qu’il a mis en place depuis la révolution tranquille. Le modèle québécois a permis le rattrapage sur l’Ontario mais, en même temps, n’a pas réussi à abaisser son taux de chômage au niveau de la moyenne canadienne. Et après mûres réflexion, l’AMEQ est d’avis que la solution ne se trouve pas dans l’adoption d’éléments issus du modèle américain. Mais, le renouveau du modèle passe par la contribution de la société civile et non seulement par des solutions étatiques, marchandes ou encore de la concertation institutionnaliséeThis short text aims at explaining the position of the Association of the manufacturers and exporters of Quebec (AMEQ on the question of the Quebecois model. For the association, there is a Quebecois model, as there is a French, American, Japanese model or other. This model, neither everything made original, nor the completely similar to the others, is liberal and social at the same moment democratic. It is characterized by the nationalism, the equity and the dialogue. The modern economic story, partially that of the role of the state by the economic control levers which it set up since the “Révolution tranquille”. The Quebecois model allowed the picking up on Ontario but, at the same time, did not manage to lower its unemployment rate at the level of the Canadian avarage. And after reflection, the AMEQ is of

Implementation of an implicit method into heat conduction calculation of TRAC-PF1/MOD2 code

International Nuclear Information System (INIS)

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

1990-08-01

A two-dimensional unsteady heat conduction equation is solved in the TRAC-PF/MOD2 code to calculate temperature transients in fuel rod. A large CPU time is often required to get stable solution of temperature transients in the TRAC calculation with a small axial node size (less than 1.0 mm), because the heat conduction equation is discretized explicitly. To eliminate the restriction of the maximum time step size by the heat conduction calculation, an implicit method for solving the heat condition equation was developed and implemented into the TRAC code. Several assessment calculations were performed with the original and modified TRAC codes. It is confirmed that the implicit method is reliable and is successfully implemented into the TRAC code through comparison with theoretical solutions and assessment calculation results. It is demonstrated that the implicit method makes the heat conduction calculation practical even for the analyses of temperature transients with the axial node size less than 0.1 mm. (author)

The Ar{sup 17+} Ly{sub {alpha}2}/Ly{sub {alpha}1} ratio in Alcator C-Mod tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Rice, J E; Reinke, M L; Ince-Cushman, A C; Podpaly, Y A [Plasma Science and Fusion Center, MIT, Cambridge, MA (United States); Ashbourn, J M A [Mathematical Institute, University of Oxford, Oxford (United Kingdom); Gu, M F [SSL, University of California Berkeley, CA (United States); Bitter, M; Hill, K [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Rachlew, E, E-mail: rice@psfc.mit.edu [KTH, Stockholm (Sweden)

2011-08-28

High-quality spectra of hydrogen-like Ar{sup 17+} have been obtained from Alcator C-Mod tokamak plasmas using a spatially imaging high-resolution x-ray spectrometer system in an extensive study of the underlying high-n satellite lines. The ratio of Ly{sub {alpha}2} (1S{sub 1/2}-2P{sub 1/2}) to Ly{sub {alpha}1} (1S{sub 1/2}-2P{sub 3/2}) was found to be {approx}0.52 regardless of plasma parameters, which is somewhat greater than the ratio of the statistical weights of the upper n = 2 levels, 0.5. This difference is mainly due to the effects of collisional excitation of fine-structure sub-levels. For the observations presented here, electron densities were in an extended range from 3x10{sup 19} to 4x10{sup 20} m{sup -3} with electron and ion temperatures between 1 and 4 keV. Experimental results are compared to calculations from COLRAD, a collisional-radiative modelling code, and good agreement is shown.

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

International Nuclear Information System (INIS)

Lozano, M.F.; Moreno, P.; de la Cal, C.; Larrea, E.; Lopez, A.; Santamaria, J.G.; Lopez, E.; Novo, M.

1993-07-01

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

Post-test analysis with RELAP5/MOD2 of ROSA-IV/LSTF natural circulation test ST-NC-02

International Nuclear Information System (INIS)

Chauliac, C.; Kukita, Yutaka; Kawaji, Masahiro; Nakamura, Hideo; Tasaka, Kanji.

1988-10-01

Results of post-test analysis for the ROSA-IV/LSTF natural circulation experiment ST-NC-02 are presented. The experiment consisted of many steady-state stages registered for different primary inventories. The calculation was done with RELAP5/MOD2 CYCLE 36.00. Discrepancies between the calculation and the experiment are observed: the core flow rate is overestimated at inventories between 80 % and 95 %; the inventory at which dryout occurs in the core is also much overestimated. The causes of these discrepancies are studies through sensitivity calculations and the following key parameters are pointed out: the interfacial friction and the form loss coefficients in the vessel riser, the SG U-tube multidimensional behaviour, the interfacial friction in the SG inlet plenum and in the pipe located underneath. (author)

Calculation study of nonequilibrium post-CHF heat transfer in rod bundle test using modified RELAP5/MOD2

International Nuclear Information System (INIS)

Hassan, Y.A.

1987-01-01

To date there is only very limited data for non-equilibrium convective film boiling in rod bundle geometries. A recent nine (3 x 3) rod bundle post-critical-flux (CHF) test from the Lehigh University test facility was simulated using RELAP5/MOD2, to assess its capabilities in predicting the overall convective mechanisms in post-CHF heat transfer in rod bundle geometries. The code calculations were compared with experimental data. The code predicted low vapor superheats and void fraction oscillations. A new interfacial heat transfer between the droplet/steam resulted in a reasonable prediction of vapor superheats. A revised dispersed flow film boiling correlation which accounts for the enhancement of steam convective cooling by droplet-induced turbulence was incorporated in the code. Comparison with the data showed a fair agreement

Experiment data report for semiscale Mod-1 test S-06-4 (LOFT counterpart test)

International Nuclear Information System (INIS)

Gillins, R.L.; Sackett, K.E.; Coppin, C.E.

1977-12-01

Recorded test data are presented for Test S-06-4 of the Semiscale Mod-1 LOFT counterpart 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-06-4 was conducted from initial conditions of 15,653 kPa and 564 K 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 broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 100 percent of the maximum peak power density

Experiment data report for semiscale Mod-1 test S-06-1 (LOFT counterpart test)

International Nuclear Information System (INIS)

Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

1977-07-01

Recorded test data are presented for Test S-06-1 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying an hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-1 was conducted from initial conditions of 15 568 kPa and 564 K 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 broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 30% of the maximum peak power density

Internal transport barriers on Alcator C-Mod

International Nuclear Information System (INIS)

Fiore, C.L.; Rice, J.E.; Bonoli, P.T.; Boivin, R.L.; Goetz, J.A.; Hubbard, A.E.; Hutchinson, I.H.; Granetz, R.S.; Greenwald, M.J.; Marmar, E.S.; Mossessian, D.; Porkolab, M.; Taylor, G.; Snipes, J.; Wolfe, S.M.; Wukitch, S.J.

2001-01-01

The formation of internal transport barriers (ITBs) has been observed in the core region of Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] under a variety of conditions. The improvement in core confinement following pellet injection (pellet enhanced performance or PEP mode) has been well documented on Alcator C-Mod in the past. Recently three new ITB phenomena have been observed which require no externally applied particle or momentum input. Short lived ITBs form spontaneously following the high confinement to low confinement mode transition and are characterized by a large increase in the global neutron production (enhanced neutron or EN modes). Experiments with ion cyclotron range of frequencies power injection to the plasma off-axis on the high field side results in the central density rising abruptly and becoming peaked. The ITB formed at this time lasts for ten energy confinement times. The central toroidal rotation velocity decreases and changes sign as the density rises. Similar spontaneous ITBs have been observed in ohmically heated H-mode plasmas. All of these ITB events have strongly peaked density profiles with a minimum in the density scale length occurring near r/a=0.5 and have improved confinement parameters in the core region of the plasma

Les industries culturelles en mutation : des modèles en question

Directory of Open Access Journals (Sweden)

Lucien Perticoz

2012-09-01

Full Text Available La présente contribution se propose de questionner la notion de modèles socio-économiques dans le cadre des travaux relatifs aux mutations des industries culturelles. À cette fin, l’exposé se déroulera en trois temps : nous reviendrons tout d’abord sur les caractéristiques essentielles des modèles génériques (modèle éditorial et de flot ainsi que sur leurs principaux apports ; nous expliquerons ensuite dans quelle mesure ils doivent être considérés, non comme une description fidèle de la réalité dont ils entendent rendre compte, mais davantage comme des règles du jeu permettant d’appréhender les mutations à l’œuvre ; enfin, à l’aune de la numérisation des contenus et de leur consommation via Internet, nous interrogerons l’hypothèse de l’émergence de nouveaux modèles génériques. En conclusion, nous insisterons sur la nécessité, à notre sens, de prendre en compte l’évolution des pratiques culturelles médiatiques en tant que dimension structurante de ces modèles.This paper aims to question the concept of socio-economic models within the framework of research about cultural industries mutations. For this purpose, our presentation will proceed in three parts : first of all, we will reconsider the essential characteristics of the generic models (publishing model and flow model and their main contributions to the research ; we will explain then why they must be considered, not as a faithful description of the reality of which they intend to give an account, but more like rules of the game allowing to understand the mutations of cultural industries ; finally, considering the digitalization of contents and their consumption using Internet, we will question the hypothesis of the emergence of new generic models. In conclusion, we will insist on the importance, from our opinion, to take into account the evolution of the media cultural practices as a structuring dimension of these models.

A comparative simulation of feed and bleed operation during the total loss of feedwater event by RELAP5/MOD3 and CEFLASH-4AS/REM computer codes

International Nuclear Information System (INIS)

Kwon, Y.M.; Ro, T.S.; Song, J.H.

1995-01-01

The Ulchin 3 and 4 nuclear power plants, which are two-loop 2,825 MW(thermal) pressurized water reactors designed by the Korea Atomic Energy Research Institute, adopted a safety depressurization system (SDS) to mitigate the beyond-design-basis event of a total loss of feedwater (TLOFW). A comparative simulation by the CEFLASH-4AS/REM and RELAP5/MOD3 computer codes for the TLOFW event without operator recovery and the TLOFW event with feed and bleed (F and B) operation is performed for Ulchin 3 and 4. In the analyses, the SDS bleed paths are modeled by orifices located on the top of the pressurizer, where the analytical area of the bleed path is based on the Ulchin 3 and 4 SDS design flow capacity. An additional case, where the SDS piping and valves are modeled explicitly, is considered for the RELAP5 analysis. The predictions by the CEFLASH-4AS/REM of the transient two-phase system behavior show good qualitative and quantitative agreement with those by the RELAP5 simulation. The RELAP5 case with explicit piping results in less repressurization and lower reactor coolant system pressure than that of the case without explicit SDS modeling. However, the two cases of RELAP5 analyses result in essentially the same transient scenarios for TLOFW with F and B operation. The results of the simulation demonstrate the validity of the Ulchin 3 and 4 design approach, which employs CEFLASH-4AS/REM computer code and SDS bleed paths modeled by orifices located on the top of the pressurizer. The results also indicate that the decay heat removal and core inventory makeup function can be successfully accomplished by F and B operation by using the SDS for Ulchin 3 and 4

Waste removal sequencing using ProdMod

International Nuclear Information System (INIS)

Paul, P.K.; Gregory, M.V.; Davis, N.R.; Brooke, J.N.

1996-01-01

The Savannah River Site (SRS) is starting to solidify its accumulated high-level radioactive waste into borosilicate glass in stainless steel canisters for eventual permanent storage. The in-tank precipitation process (ITP) and extended sludge processing (ESP) are two key operations in the waste processing complex. The supernate and dissolved salt from the waste storage tanks are transferred to the ITP process tank where the solution is decontaminated in batch processes. Soluble radioactive cesium is precipitated with sodium tetraphenylborate and strontium, uranium, and plutonium are adsorbed on monosodium titanate. The precipitate and adsorbent solids, which now contain the radionuclides, are concentrated using crossflow filters. The concentrated solids are sent to the high-level waste vitrification process. The decontaminated salt solution is sent to the low-level waste solidification process to form cement grout. In parallel with the precipitate operations, insoluble sludges that settled originally to the bottom of the waste tanks are reslurried and sent to ESP to undergo washing to reduce soluble salt content and aluminum dissolution, if required. In the vitrification process in the Defense Waste Processing Facility (DWPF), the concentrated precipitate from the ITP is mixed with the washed sludge from ESP and glass frit in proportion to form a stable borosilicate glass. A novel and fast-running Production Planning Model (ProdMod) has been developed to simulate the waste processing operation. This paper describes the application of ProdMod in sequencing the ITP batches and scheduling the ESP batches

Mechanical measurements in RFX-mod experiment

Energy Technology Data Exchange (ETDEWEB)

Dalla Palma, M., E-mail: mauro.dallapalma@igi.cnr.it [Consorzio RFX, EURATOM-ENEA Association, Corso Stati Uniti 4, 35127 Padova (Italy); Ravarotto, D.; Dal Bello, S.; Fincato, M.; Ghiraldelli, R.; Marchiori, G.; Taliercio, C.; Zaccaria, P. [Consorzio RFX, EURATOM-ENEA Association, Corso Stati Uniti 4, 35127 Padova (Italy)

2010-12-15

The ongoing experimental campaigns of RFX-mod are performed setting operational parameters at the nominal or exceeding design values of the experiment. Compressive forces up to 11 MN are produced by high magnetizing currents that reach values up to 50 kA. These forces heavily load the mechanical structure of RFX-mod and their effects are measured and monitored in order to assess the expected structural response and machine reliability during operation. Mechanical strains and relative displacements are real time measured during each experimental pulse by 48 strain gauges located on 12 mechanical struts and 16 potentiometers positioned between the toroidal assembly and the support mechanical structure. The strains in the most stressed components are measured by means of 24 half bridge gauges preliminarily calibrated. Particular care has been given to sensor choices and installation in order to minimize signal noises induced by the electrical and magnetic fields. The residual noises have been further reduced by proper sampling frequency and averaging techniques. The strains measured on the struts are then post-processed to calculate the resultant forces and bending moments, while the displacement measurements give an estimate of the overall stiffness of the mechanical structure. The measured forces and displacements are shown per toroidal locations and as a function of the current intensity, so verifying the uniform sharing of forces among the different struts and the proper square law correlation with the magnetizing current intensity.

Analyse géographique et modélisation des dynamiques d’urbanisation à La Réunion

Directory of Open Access Journals (Sweden)

Pascal Thinon

2007-07-01

Full Text Available Cet article propose un prototype de modèle d’interprétation des dynamiques urbaines sur l’Ile de la Réunion. En entrée, le modèle combine un ensemble de champs géographiques jouant en faveur ou en défaveur de l’urbanisation ; en sortie, il indique une propension à l’urbanisation de chaque lieu. Il est conçu de manière à permettre une analyse exploratoire de ces dynamiques selon une approche heuristique. Les premiers résultats ont permis d’obtenir, sur l’ensemble de l’île, une carte de la propension à l’urbanisation, jugée satisfaisante eu égard aux dynamiques observées entre 1989 et 2002 ; un premier scénario d’évolution, concernant les espaces de savanes à l’ouest de l’île, est également proposé. Les premiers travaux sur ce modèle, encore à un stade préliminaire, sont encourageants mais soulèvent de nombreuses questions concernant notamment le calibrage des facteurs et leur rôle respectif, l’intégration de nouveaux champs comme le voisinage, l’analyse des résidus, la mesure de la qualité du modèle ou bien encore la mobilisation de ce type de modèle comme outil d’accompagnement de projets de territoires.

Investigation of RF-enhanced plasma potentials on Alcator C-Mod

Energy Technology Data Exchange (ETDEWEB)

Ochoukov, R., E-mail: ochoukov@psfc.mit.edu [PSFC MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States); Whyte, D.G.; Brunner, D. [PSFC MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States); Cziegler, I. [Center for Energy Research, UCSD, 9500 Gilman Drive, La Jolla, CA 92093 (United States); LaBombard, B.; Lipschultz, B. [PSFC MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States); Myra, J. [Lodestar Research Corporation, 2400 Central Avenue P-5, Boulder, CO 80301 (United States); Terry, J.; Wukitch, S. [PSFC MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States)

2013-07-15

Radio frequency (RF) sheath rectification is a leading mechanism suspected of causing anomalously high erosion of plasma facing materials in RF-heated plasmas on Alcator C-Mod. An extensive experimental survey of the plasma potential (Φ{sub P}) in RF-heated discharges on C-Mod reveals that significant Φ{sub P} enhancement (>100 V) is found on outboard limiter surfaces, both mapped and not mapped to active RF antennas. Surfaces that magnetically map to active RF antennas show Φ{sub P} enhancement that is, in part, consistent with the recently proposed slow wave rectification mechanism. Surfaces that do not map to active RF antennas also experience significant Φ{sub P} enhancement, which strongly correlates with the local fast wave intensity. In this case, fast wave rectification is a leading candidate mechanism responsible for the observed enhancement.

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

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

Primes of the form x2 + dy2 with x ≡ 0(mod N)

Indian Academy of Sciences (India)

Department of Computer Science and Automation, Indian Institute of Science, ... This led many mathematicians to work on primes of the form x2 + dy2 for d = 2, 3, 5, 7 ..... Chandan Singh Dalawat, H.R.I. Allahabad, for the kind help and.

C-Mod Collaboration Informal Technical Progress Report

International Nuclear Information System (INIS)

Kenneth W. Gentle

2007-01-01

The aims of the collaboration have not changed. A specific list of tasks was agreed upon during the Fall of 2006 in preparation for the 2007 C-Mod campaign by Earl Marmar, Head of the Alcator Project, Kenneth Gentle, Principal Investigator, and William Rowan, Collaboration Coordinator with the facilitation of Adam Rosenberg (DOE grant monitor for the collaboration). The activities follow the list of tasks and are discussed in this progress report

Intelligence artificielle et agents collectifs : le modèle EUROSIM

Directory of Open Access Journals (Sweden)

Denise Pumain

2007-07-01

Full Text Available EUROSIM est un modèle multi-agents conçu pour simuler l’évolution à moyen terme du système des villes européennes. Les agents sont des entités collectives, les grandes villes caractérisées par leur taille et leur fonction dans le système des villes, et dont les interactions (échanges modulés par des relations de proximité ou de réseau déterminent la dynamique relative, tandis que la croissance d’ensemble dépend de l’innovation. Des outils d’analyse multiscalaire ont été développés afin d’interpréter les sorties du modèle et faciliter le calibrage.

TRAC code development status and plans

International Nuclear Information System (INIS)

Spore, J.W.; Liles, D.R.; Nelson, R.A.

1986-01-01

This report summarizes the characteristics and current status of the TRAC-PF1/MOD1 computer code. Recent error corrections and user-convenience features are described, and several user enhancements are identified. Current plans for the release of the TRAC-PF1/MOD2 computer code and some preliminary MOD2 results are presented. This new version of the TRAC code implements stability-enhancing two-step numerics into the 3-D vessel, using partial vectorization to obtain a code that has run 400% faster than the MOD1 code

Performance Improvement of Real-Time System for Plasma Control in RFX-mod

International Nuclear Information System (INIS)

Luchetta, A.; Manduchi, G.; Soppelsa, A.; Taliercio, C.

2006-01-01

The real-time system for plasma control has been used routinely in RFX-mod since commissioning (mid 2005). It is based on a modular hardware/software infrastructure, currently including 7 VME stations, capable of fulfilling the tight system requirements in terms of input/output channels (> 700 / > 250), real-time data flow (> 2 Mbyte/s), computation capability (> 1 GFLOP/s per station), and real-time constraints (application cycle times rd EPS Conf. on Plasma Physics, Rome Italy, June 19 - 23 2006]. The high flexibility of the system has stimulated the development of a large number of control schemes with progressively increasing requests in terms of computation complexity and real-time data flow, demanding, at the same time, strict control on cycle times and system latency. Even though careful optimisation of algorithm implementation and real-time data transmission have been performed, allowing to keep pace, so far, with the increased control requirements, future developments require to evolve the current technology, retaining the basic architecture and concepts. Two system enhancements are envisaged in the near future. The 500 MHz PowerPC-based Single Board Computer currently in use will be substituted with the 1 GHz version, whereas the real-time communication system will increase in bandwidth from 100 Mbit/s to 1 Gbit/s. These improvements will surely enhance the overall system performance, even if it is not possible to quantify a priori the exact performance boost, since other components may limit the performance in the new configuration. The paper reports in detail on the analysis of the bottlenecks of the current architecture. Based on measurements carried out in laboratory, it presents the results achieved with the proposed enhancements in terms of real-time data throughput, cycle times and latency. The paper analyses in detail the effects of the increased computing power on the components of the control system and of the improved bandwidth in real

Overview of Alcator C-Mod Research

Science.gov (United States)

White, A. E.

2017-10-01

Alcator C-Mod, a compact (R =0.68m, a =0.21m), high magnetic field, Bt Research spans the topics of core transport and turbulence, RF heating and current drive, pedestal physics, scrape-off layer, divertor and plasma wall interactions. In the last experimental campaign, Super H-mode was explored and featured the highest pedestal pressures ever recorded, pped 90 kPa (90% of ITER target), consistent with EPED predictions. Optimization of naturally ELM-suppressed EDA H-modes accessed the highest volume averaged pressures ever achieved (〈p〉>2 atm), with pped 60 kPa. The SOL heat flux width has been measured at Bpol = 1.25T, confirming the Eich scaling over a broader poloidal field range than before. Multi-channel transport studies focus on the relationship between momentum transport and heat transport with perturbative experiments and new multi-scale gyrokinetic simulation validation techniques were developed. U.S. Department of Energy Grant No. DE-FC02-99ER54512.

Models for the Behavior of Offshore Structure Foundations. Part One: Methodologies and Rheological Models for Soils Modèles pour le comportement des fondations d'ouvrages types marins. Première partie : Méthodologies et modèles rhéologiques de sols

Directory of Open Access Journals (Sweden)

Meimon Y.

2006-11-01

structure. Prospects opened up by the research are discussed. La conception des fondations de grands ouvrages est une tâche complexe qui requiert à la fois l'expérience de l'expert et l'utilisation de modèles numériques adéquats pour assurer la sécurité et optimiser les coûts de dimensionnement. En fait, prévoir le comportement d'une fondation nécessite de bien évaluer les effets combinés de la technique de mise en place, de la variabilité spatiale des propriétés mécaniques, de l'incertitude sur les chargements et des techniques de modélisation du comportement mécanique des géomatériaux. Ceci est particulièrement vrai pour les plates-formes marines, qu'elles soient destinées à l'exploration ou à la production du pétrole, dans la mesure où les chargements non-monotones dus à l'environnement marin, souvent très sévère, peuvent avoir des effets très néfastes sur le comportement de ces structures. On présente, en deux parties, la synthèse d'une dizaine d'années d'activités de recherche, menées par une équipe de l'Institut Français du Pétrole (IFP, en collaboration avec plusieurs équipes universitaires et des centres techniques et industriels, pour la mise au point de méthodologies et d'outils adaptés au calcul du comportement des fondations d'ouvrages types marins durant toute la durée de vie de la plate-forme. Cet article concerne la première partie qui est dévolue à l'exposé des démarches développées et à la modélisation de l'état local des géomatériaux par des modèles rhéologiques. Démarches de modélisation du comportement des fondations d'ouvrages marins : On analyse d'abord la spécificité des fondations de plates-formes marines en insistant sur les différences avec les ouvrages à terre : dimension inusitée, chargement aléatoire par les éléments marins. On peut alors dégager plusieurs types de fondation (fig. 2. 1 et 2. 2 et les classes de problèmes qu'il est important d'étudier, qui recouvrent aussi

Modelling of advanced tokamak physics scenarios in ALCATOR C-Mod

International Nuclear Information System (INIS)

Bonoli, P.T.; Porkolab, M.; Ramos, J.

2001-01-01

Advanced tokamak modes of operation in Alcator C-Mod have been investigated using a simulation model which combines an MHD equilibrium and current profile control calculation with an ideal MHD stability analysis. Stable access to high β t operating modes with reversed shear current density profiles has been demonstrated using 2.4-3.0 MW of off-axis lower hybrid current drive (LHCD). Here β t =2μ 0 (p)/B 2 0 is the volume averaged toroidal plasma beta. Current profile control at the β-limit and beyond has also been demonstrated. The effects of LH power level as well as changes in the profiles of density and temperature on shear reversal radius have been quantified and are discussed. (author)

RELAP5/MOD 3.2 Analysis of the Loss of RHR System Experiment Scaled to NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Bajs, T.; Prah, M.

1998-01-01

In the paper the RELAP5/MOD 3.2 analysis of the loss of Residual Heat Removal (RHR) system during midloop operation experiment performed at the Rig of Safety Assessment (ROSA)-IV/Large Scale Test Facility (LSTF) together with the analysis of the same test scenario scaled to NPP Krsko are presented. The experiment consisted in a loss of the RHR system at cold shutdown conditions along with a 5% cold leg break in the loop without pressurizer. The Safety Injection (SI) system was disable in the calculation. The aims of the work were to study the physical phenomena encountered under low power and low system pressure conditions while the upper part of the Reactor Coolant System (RCS) is filled with noncondensable. The impact of the bypass flow between upper plenum and downcomer inlet on transient responses was investigated. The transient was simulated for 6000 s. (author)

Verification of LOCA/ECCS analysis codes ALARM-B2 and THYDE-B1 by comparison with RELAP4/MOD6/U4/J3

International Nuclear Information System (INIS)

Shimizu, Takashi

1982-08-01

For a verification study of ALARM-B2 code and THYDE-B1 code which are the component of the JAERI code system for evaluation of BWR ECCS performance, calculations for typical small and large break LOCA in BWR were done, and compared with those by RELAP4/MOD6/U4/J3 code. This report describes the influences of differences between the analytical models incorporated in the individual code and the problems identified by this verification study. (author)

Mod-5A wind turbine generator program design report. Volume 4: Drawings and specifications, book 5

Science.gov (United States)

1984-01-01

The design, development and analysis of the 7.3 MW MOD-5A wind turbine generator is documented. There are four volumes. This volume contains the drawings and specifications that were developed in preparation for building the MOD-5A wind turbine generator. Detail drawings of several assemblies and subassemblies are given. This is the fifth book of volume 4.

Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test)

International Nuclear Information System (INIS)

1977-06-01

Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart 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-06-5 was conducted from initial conditions of 2272 psia and 536 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 broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations

Comparison of tungsten nano-tendrils grown in Alcator C-Mod and linear plasma devices

International Nuclear Information System (INIS)

Wright, G.M.; Brunner, D.; Baldwin, M.J.; Bystrov, K.; Doerner, R.P.; Labombard, B.; Lipschultz, B.; De Temmerman, G.; Terry, J.L.; Whyte, D.G.; Woller, K.B.

2013-01-01

Growth of tungsten nano-tendrils (“fuzz”) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The depth of the W fuzz layer (600 ± 150 nm) is consistent with an empirical growth formula from the PISCES experiment. Re-creating the C-Mod exposures as closely as possible in Pilot-PSI experiment can produce nearly-identical nano-tendril morphology and layer thickness at surface temperatures that agree with uncertainties with the C-Mod W probe temperature data. Helium concentrations in W fuzz layers are measured at 1–4 at.%, which is lower than expected for the observed sub-surface voids to be filled with several GPa of helium pressure. This possibly indicates that the void formation is not pressure driven

Modèle de compaction élasto-plastique en simulation de bassins Elastoplastic Compaction Model for Basin Simulation

Directory of Open Access Journals (Sweden)

Schneider F.

2006-11-01

stress is close to 1, which is coherent with the hypothesis of considering the grains making up the skeleton as being indeformable. COMP1D is a software that makes a 1D simulation of the geologic history of a sedimentary column. It integrates the physical phenomena described in the first part. The geometric variations caused by sedimentation, erosions and compaction are taken into consideration by introducing various Lagrangian coordinate systems (Fig. 2. The temperature is imposed by a surface temperature and a gradient. The boundary conditions are the ones conventionally used in basin simulators. There are various versions of the software corresponding to different numerical approaches. Our problem was discretized by finite-element methods with linear shape functions, finite-element methods with quadratic shape functions, and finite-volume methods. Numerous tests showed that the pressure solution to convergence is identical for all such methods. However finite-element methods cannot be used to compute a velocity field for the fluid that gives perfect local conservation. This local conservation is absolutely necessary for coupling a transport equation (heat equation, saturation equation for two-phase flows with the computing of the pressure of the fluid. Only finite-volume methods, which handle nonlinearities (Newton's method correctly, are locally perfectly conservative. This model improves preceding models, mainly by introducing the concept of elasticity. However, as things now stand and from the theoretical standpoint, it is valid only in the superficial layer of sediments. Its extension to deeper layers is acceptable in basin simulators only because the porosity variation is slight. The introduction of an alpha coefficient in the definition of the effective stress seems necessary. However, the fact of taking this coefficient different from 1 must, for reasons of coherence, lead us to give increased consideration to the mechanical deformations of the grains. The

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

Edge Zonal Flows and Blob Propagation in Alcator C-Mod

International Nuclear Information System (INIS)

Zweben, S.; Terry, J.L.; Agostini, M.; Davis, B.; Grulke, O.; Hager, R.; Hughes, J.; LaBombard, B.; D'Ippolito, D.A.; Myra, J.R.; Russell, D.A.

2011-01-01

Here we describe recent measurements of the 2-D motion of turbulence in the edge and scrape-off layer (SOL) of the Alcator C-Mod tokamak. This data was taken using the outer midplane gas puff imaging (GPI) camera, which views a 6 cm radial by 6 cm poloidal region near the separatrix just below the outer midplane [1]. The data were taken in Ohmic or RF heated L-mode plasmas at 400,000 frames/sec for ∼50 msec/shot using a Phantom 710 camera in a 64 x 64 pixel format. The resulting 2-D vs. time movies [2] can resolve the structure and motion of the turbulence on a spatial scale covering 0.3-6 cm. The images were analyzed using either a 2-D cross-correlation code (Sec. 2) or a 2-D blob tracking code (Sec. 3).