Implementation of the thermal-hydraulic transient analysis code RELAP4/MOD5 and MOD6 on the FACOM 230/75 computer system

International Nuclear Information System (INIS)

Kohsaka, Atsuo; Ishigai, Takahiro; Kumakura, Toshimasa; Naraoka, Ken-itsu

1979-03-01

Development efforts have continued on the extensively used LOCA analysis code RELAP-4, as seen in its history; that is, from the prototype version MOD2 to the latest one MOD6 which is capable of one-through calculations from blowdown to reflood phase of PWR-LOCA. Many improvements and refinements of the models have enlarged the scopes and extents of phenomena to treat. Correspondingly the size of program has increased version to version, and special programming techniques have continuously been introduced to manage the program within limited capacity of core memory. For example, the Dynamic Storage Allocation of MOD5 and the PRELOAD Preprocessor newly incorporated in MOD6 are those designed for the CDC computer with relatively small core size. Described are these programming techniques in detail and experiences on implementation of the codes on FACOM 230/75, together with some results of confirmatory calculations. (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)

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)

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)

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)

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

Combining structural-thermal coupled field FE analysis and the Taguchi method to evaluate the relative contributions of multi-factors in a premolar adhesive MOD restoration.

Science.gov (United States)

Lin, Chun-Li; Chang, Yen-Hsiang; Lin, Yi-Feng

2008-08-01

The aim of this study was to determine the relative contribution of changes in restorative material, cavity dimensions, adhesive layer adaptation, and load conditions on the biomechanical response of an adhesive Class II MOD restoration during oral temperature changes. A validated finite-element (FE) model was used to perform the structural-thermal coupled field analyses and the Taguchi method was employed to identify the significance of each design factor in controlling the stress. The results indicated that thermal expansion in restorative material amplified the thermal effect and dominated the tooth stress value (69%) at high temperatures. The percentage contributions of the load conditions, cavity depth, and cement modulus increased the effect on tooth stress values 46%, 32%, and 14%, respectively, when the tooth temperature was returned to 37 degrees C. Load conditions were also the main factor influencing the resin cement stress values, irrespective of temperature changes. Increased stress values occurred with composite resin, lateral force, a deeper cavity, and a higher luting cement modulus. The combined use of FE analysis and the Taguchi method efficiently identified that a deeper cavity might increase the risk of a restored tooth fracture, as well as a ceramic inlay with a lower thermal expansion, attaining a proper occlusal adjustment to reduce the lateral occlusal force and low modulus luting material application to obtain a better force-transmission mechanism are recommended.

Studies of radiant heat transfer problems by the MOXY-program

International Nuclear Information System (INIS)

Wennerberg, D.; Thiede, M.

1988-01-01

MOXY is a program for calculation of transients at LOCA in a BWR. The program has been enlarged for application to 9 x 9 bundles (earlier only 7 x 7 - and 8 x 8 geometries). The report presents the results of five runs, two cases for 8 x 8 -bundle and three for 9 x 9 bundle. Comparison is made with estimates made by other, similar programs. (O.S.)

An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations

International Nuclear Information System (INIS)

Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E.; Nikulshin, V.

1996-09-01

This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases

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

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.

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.

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)

Core thermal response during Semiscale Mod-1 blowdown heat transfer tests

International Nuclear Information System (INIS)

Larson, T.K.

1976-06-01

Selected experimental data and results calculated from experimental data obtained from the Semiscale Mod-1 PWR blowdown heat transfer test series are analyzed. These tests were designed primarily to provide information on the core thermal response to a loss-of-coolant accident. The data are analyzed to determine the effect of core flow on the heater rod thermal response. The data are also analyzed to determine the effects of initial operating conditions on the rod cladding temperature behavior during the transient. The departure from nucleate boiling and rewetting characteristics of the rod surfaces are examined for radial and axial patterns in the response. Repeatability of core thermal response data is also investigated. The test data and the core thermal response calculated with the RELAP4 code are compared

Thermal fatigue crack growth tests and analyses of thick wall cylinder made of Mod.9Cr–1Mo steel

Energy Technology Data Exchange (ETDEWEB)

Wakai, Takashi, E-mail: wakai.takashi@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita-cho Oarai, Ibaraki 3111393 (Japan); Inoue, Osamu [IX Knowledge Inc., 3-22-23 MSC Center Bldg, Kaigan Minato-ku, Tokyo 1080022 Japan (Japan); Ando, Masanori; Kobayashi, Sumio [Japan Atomic Energy Agency, 4002 Narita-cho Oarai, Ibaraki 3111393 (Japan)

2015-12-15

Highlights: • A thermal fatigue crack growth test was performed using Mod.9Cr–1Mo steel cylinder. • Axial/circumferential notches were machined on the inner surface of the cylinder. • Simplified analytical results were compared to the test data. • Crack length could not be predicted by the analyses because of crack conjunctions. • If there are no surface cracks, the calculations might agree with the observations. - Abstract: In Japan, the basic designing works for a demonstration plant of Japan Sodium cooled Fast Reactor (JSFR) are now conducted. JSFR is an advanced loop type reactor concept. To enhance the safety and the economic competitiveness, JSFR employs modified 9% chromium–1% molybdenum (Mod.9Cr–1Mo) steel as a material for coolant pipes and components, because the steel has both excellent high temperature strength and thermal properties. The steel has been standardized as a nuclear material in Japan Society of Mechanical Engineers (JSME) code in 2012. In JSFR pipes, demonstration of Leak Before Break (LBB) aspect is strongly expected because the safety assessment may be performed on the premise of leak rate where the LBB aspect is assured. Although the authors have already performed a series of thermal fatigue crack growth tests of austenitic stainless steel cylinders (Wakai et al., 2005), crack growth behavior in the structures made of Mod.9Cr–1Mo steel has not been investigated yet. Especially for the welded joints of Mod.9Cr–1Mo steel, “Type-IV” cracking may occur at heat affected zone (HAZ). Therefore, this study performed a series of thermal fatigue crack growth tests of thick wall cylinders made of Mod.9Cr–1Mo steel including welds, to obtain the crack growth data under cyclic thermal transients. The test results were compared to the analytical results obtained from JAEA's simplified methods (Wakai et al., 2005).

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)

Analysis of loss of normal feedwater transient using RELAP5/MOD1/NSC; KNU1 plant simulation

International Nuclear Information System (INIS)

Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo

1986-01-01

Simulation of the system thermal-hydraulic parameters was carried out following the KNU1(Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on november 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS(Reactor Coolant System) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018. (Author)

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)

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

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

MoXy fiber with active cooling cap for bovine prostate vaporization with high power 200W 532 nm laser

Science.gov (United States)

Peng, Steven Y.; Kang, Hyun Wook; Pirzadeh, Homa; Stinson, Douglas

2011-03-01

A novel MoXyTM fiber delivery device with Active Cooling Cap (ACCTM) is designed to transmit up to 180W of 532 nm laser light to treat benign prostatic hyperplasia (BPH). Under such high power tissue ablation, effective cooling is key to maintaining fiber power transmission and ensuring the reliability of the fiber delivery device To handle high power and reduce fiber degradation, the MoXy fiber features a larger core size (750 micrometer) and an internal fluid channel to ensure better cooling of the fiber tip to prevent the cap from burning, detaching, or shattering during the BPH treatment. The internal cooling channel was created with a metal cap and tubing that surrounds the optical fiber. In this study MoXy fibers were used to investigate the effect of power levels of 120 and 200 W on in-vitro bovine prostate ablation using a 532 nm XPSTM laser system. For procedures requiring more than 100 kJ, the MoXy fiber at 200W removed tissue at twice the rate of the current HPS fiber at 120W. The fiber maintained a constant tissue vaporization rate during the entire tissue ablation process. The coagulation at 200W was about 20% thicker than at 120W. In conclusion, the new fibers at 200W doubled the tissue removal rate, maintained vaporization efficiency throughout delivery of 400kJ energy, and induced similar coagulation to the existing HPS fiber at 120W.

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

TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

International Nuclear Information System (INIS)

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

1986-07-01

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients

TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

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

1986-07-01

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

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)

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)

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.

Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

Energy Technology Data Exchange (ETDEWEB)

No, Hee Cheon; Park, Hyun Sik; Kim, Hyoung Tae; Moon, Young Min; Choi, Sung Won; Hwang, Do Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

2000-03-15

The direct-contact condensation hear transfer coefficients are experimentally obtained in the following conditions : pure steam/steam in the presence of noncondensible gas, horizontal/slightly inclined pipe, cocurrent/countercurrent stratified flow with water. The empirical correlation for liquid Nusselt number is developed in conditions of the slightly inclined pipe and the cocurrent stratified flow. The several models - the wall friction coefficient, the interfacial friction coefficient, the correlation of direct-contact condensation with noncondensible gases, and the correlation of wall film condensation - in the RELAP5/MOD3.2 code are modified, As results, RELAP5/MOD3.2 is improved. The present experimental data is used for evaluating the improved code. The standard RELAP5/MOD3.2 code is modified using the non-iterative modeling, which is a mechanistic model and does not require any interfacial information such as the interfacial temperature, The modified RELAP5/MOD3.2 code os used to simulate the horizontally stratified in-tube condensation experiment which represents the direct-contact condensation phenomena in a hot leg of a nuclear reactor. The modeling capabilities of the modified code as well as the standard code are assessed using several hot-leg condensation experiments. The modified code gives better prediction over local experimental data of liquid void fraction and interfacial heat transfer coefficient than the standard code. For the separate effect test of the thermal-hydraulic phenomena in the pressurizer, the scaling analysis is performed to obtain a similarity of the phenomena between the Korea Standard Nuclear Power Plant(KSNPP) and the present experimental facility. The diameters and lengths of the hot-leg, the surge line and the pressurizer are scaled down with the similitude of CCFL and velocity. The ratio of gas flow rate is 1/25. The experimental facility is composed of the air-water supply tank, the horizontal pipe, the surge line and the

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

RELAP5/MOD2 code assessment

International Nuclear Information System (INIS)

Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

1985-01-01

Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G

RELAP5/MOD2 code assessment

Energy Technology Data Exchange (ETDEWEB)

Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

1985-11-01

Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G.

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)

Thermal Aging Evaluation of Mod. 9Cr-1Mo Steel using Nonlinear Rayleigh Waves

Energy Technology Data Exchange (ETDEWEB)

Joo, Young-Sang; Kim, Hoe-Woong; Kim, Jong-Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Marino, Daniel; Kim, Jin-Yeon; Jacobs, L.J [Georgia Institute of Technology, Atlanta (United States); Ruiz, Alberto [UMSNH, Morelia (Mexico)

2014-10-15

Thermal aging can pose a high risk to decreases in the mechanical properties such as strength or creep resistance. This can lead to an unexpected failure during long term operation. Nonlinear NDE techniques are preferred over conventional NDE techniques (linear ultrasonic measurements) because nonlinear ultrasonic techniques have shown their capability to detect a microstructural damage in the structures undergoing fatigue and creep. These nonlinear ultrasonic techniques make use of the fact that the dislocation density increases, which will create a nonlinear distortion of an ultrasonic wave; this damage causes the generation of measurable higher harmonic components in an initially mono-chromatic ultrasonic signal. This study investigates the recently developed non-contact nonlinear ultrasonic technique to detect the microstructural damage of mod. 9Cr-1Mo steel based on nonlinear Rayleigh wave with varying propagation distances. Nonlinear Rayleigh surface wave measurements using a non-contact, air-coupled ultrasonic transducer have been applied for the thermal aging evaluation of modified 9Cr-1Mo ferritic-martensitic steel. Thermal aging for various heat treatment times of mod.. 9Cr-1Mo steel specimens is performed to obtain the nucleation and growth of precipitated particles in specimens. The amplitudes of the first and second harmonics are measured along the propagation distance and the relative nonlinearity parameter is obtained from these amplitudes. The relative nonlinearity parameter shows a similar trend with the Rockwell C hardness.

Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2

International Nuclear Information System (INIS)

Ojeda S, J.; Morales S, J.; Chavez M, C.

2009-10-01

This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

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

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

Assessment of RELAP5/Mod3 system thermal hydraulic code using power test data of a BWR6 reactor

International Nuclear Information System (INIS)

Lee, M.; Chiang, C.S.

1997-01-01

The power test data of Kuosheng Nuclear Power Plant were used to assess RELAP5/Mod3 system thermal hydraulic analysis code. The plant employed a General Electric designed Boiling Water Reactor (BWR6) with rated power of 2894 MWth. The purpose of the assessment is to verify the validity of the plant specific RELAP5/Mod3 input deck for transient analysis. The power tests considered in the assessment were 100% power generator load rejection, the closure of main steam isolation valves (MSIVs) at 96% power, and the trip of recirculation pumps at 68% power. The major parameters compared in the assessment were steam dome pressure, steam flow rate, core flow rate, and downcomer water level. The comparisons of the system responses predicted by the code and the power test data were reasonable which demonstrated the capabilities of the code and the validity of the input deck. However, it was also identified that the separator model of the code may cause energy imbalance problem in the transient calculation. In the assessment, the steam separators were modeled using time-dependent junctions. In the approach, a complete separation of steam and water was predicted. The system responses predicted by RELAP5/Mod3 code were also compared with those from the calculations of RETRAN code. When these results were compared with the power test data, the predictions of the RETRAN code were better than those of RELAP5/Mod3. In the simulation of 100% power generator load rejection, it was believed that the difference in the steam separator model of these two codes was one of the reason of the difference in the prediction of power test data. The predictions of RELAP/Mod3 code can also be improved by the incorporation of one-dimensional kinetic model. There was also some margin for the improvement of the input related to the feedwater control system. (author)

Analysis of the VVER Standard Problem INSC-PSBV1 '11% Coolant Leak from Upper Plenum' with RELAP5/MOD3.2

International Nuclear Information System (INIS)

Melikhov, O.; Melikhov, V.; Parfenov, Yu.; Gavritenkova, O.; Lipatov, I.; Elkin, I.; Bayless, P.

2004-01-01

Analyses of a loss-of-coolant experiment carried out at the PSB-VVER test facility with the RELAP5/MOD3.2 code have been performed independently by analysts at the Electrogorsk Research and Engineering Center (EREC) and the Idaho National Engineering and Environmental Laboratory (INEEL). The PSB-VVER facility is a full-height scale model of a VVER 1000 reactor that is approximately 1/300 scale in volume and power. VVER Standard Problem INSC-PSBV1 represents an 11% leak from the upper plenum of the PSB-VVER facility, simulating the rupture of one of the accumulator injection lines. The safety-significant thermalhydraulic phenomena occurring in VVER type reactors addressed by this experiment were identified in the test validation matrix. Most of the phenomena of the validation matrix were reasonably simulated by RELAP5/MOD3.2 in both calculations. The major differences between the test and the calculations were the timing of the core heatup, and the thermal response to the accumulator injection cycles in both calculations. The INEEL calculation had a more extensive axial heatup, with most of the core experiencing small heat-ups. The accumulator injection was more effective in quenching the core in the test than in the INEEL calculation. This difference is attributed to the liquid distribution in the core, rather than to the heat transfer models in the code. The code calculation had a more uniform axial distribution of the liquid in the core, and the accumulator injection did not have much impact on the core liquid inventory. In the EREC calculation, only one heatup of the cladding temperature was observed for upper and middle section of the fuel rods before the final heatup. The small heat-ups were not reproduced in EREC calculation. The difference could be attributed to differences in liquid distribution, namely the core region in the EREC calculation contains more liquid over most of the transient than in the experiment. The distribution of liquid in the core in

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)

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.

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

Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

International Nuclear Information System (INIS)

Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

2008-01-01

The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

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)

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

13C-Tracer Experiments in DIII-D Preliminary to Thermal Oxidation Experiments to Understand Tritium Recovery in DIII-D, JET, C-Mod, and MAST

International Nuclear Information System (INIS)

Stangeby, P.; Allen, S.; Bekris, N.; Brooks, N.; Christie, K.; Chrobak, C.; Coad, J.; Counsell, G.; Davis, J.; Elder, J.; Fenstermacher, M.; Groth, M.; Haasz, A.; Likonen, J.; Lipschultz, B.; McLean, A.; Philipps, V.; Porter, G.; Rudakov, D.; Shea, J.; Wampler, W.; Watkins, J.; West, W.; Whyte, D.

2006-01-01

Retention of tritium in carbon co-deposits is a serious concern for ITER. Developing a reliable in-situ removal method of the co-deposited tritium would allow the use of carbon plasma-facing components which have proven reliable in high heat flux conditions and compatible with high performance plasmas. Thermal oxidation is a potential solution, capable of reaching even hidden locations. It is necessary to establish the least severe conditions to achieve adequate tritium recovery, minimizing damage and reconditioning time. The first step in this multi-machine project is 13 C-tracer experiments in DIII-D, JET, C-Mod and MAST. In DIII-D and JET, 13 CH 4 has been (and in C-Mod and MAST, will be) injected toroidally symmetrically, facilitating quantification and interpretation of the results. Tiles have been removed, analyzed for 13 C content and will next be evaluated in a thermal oxidation test facility in Toronto with regard to the ability of different severities of oxidation exposure to remove the different types of (known and measured) 13 C co-deposit. Removal of D/T from B on Mo tiles from C-Mod will also be tested. OEDGE interpretive code analysis of the 13 C deposition patterns is used to generate the understanding needed to apply findings to ITER. First results are reported here for the 13 C injection experiments IN DIII-D

Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

Energy Technology Data Exchange (ETDEWEB)

Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2000-03-15

The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.

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

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

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

Analysis of Reactor Pressurized Thermal Shock Conditions Considering Upgrading of Systems Important to Safety

International Nuclear Information System (INIS)

Mazurok, A.S; Vyshemirskyij, M.P.

2015-01-01

The paper analyzes conditions of pressurized thermal shock on the reactor pressure vessel taking into account upgrading of the emergency core cooling system and primary overpressure protection system. For representative accident scenarios, calculation and comparative analysis was carried out. These scenarios include a small leak from the hot leg and PRZ SV stuck opening with re closure after 3600 sec and 3 SG heat transfer tube rupture. The efficiency of mass flow control by valves on the pump head (emergency core cooling systems) and cold overpressure protection (primary overpressure protection system) was analyzed. The thermal hydraulic model for RELAP5/Mod3.2 code with detailed downcomer (DC) model and changes in accordance with upgrades was used for calculations. Detailed (realistic) modeling of piping and equipment was performed. The upgrades prevent excessive primary cooling and, consequently, help to preserve the RPV integrity and to avoid the formation of a through crack, which can lead to a severe accident

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

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

Thermal enhancement cartridge heater modified (TECH Mod) tritium hydride bed development, Part 1 - Design and fabrication

Energy Technology Data Exchange (ETDEWEB)

Klein, J.E.; Estochen, E.G. [Savannah River National Laboratory, Aiken, SC (United States)

2015-03-15

The Savannah River Site (SRS) tritium facilities have used first generation (Gen1) LaNi{sub 4.25}Al{sub 0.75} (LANA0.75) metal hydride storage beds for tritium absorption, storage, and desorption. The Gen1 design utilizes hot and cold nitrogen supplies to thermally cycle these beds. Second and third generation (Gen2 and Gen3) storage bed designs include heat conducting foam and divider plates to spatially fix the hydride within the bed. For thermal cycling, the Gen2 and Gen3 beds utilize internal electric heaters and glovebox atmosphere flow over the bed inside the bed external jacket for cooling. The currently installed Gen1 beds require replacement due to tritium aging effects on the LANA0.75 material, and cannot be replaced with Gen2 or Gen3 beds due to different designs of these beds. At the end of service life, Gen1 bed desorption efficiencies are limited by the upper temperature of hot nitrogen supply. To increase end-of-life desorption efficiency, the Gen1 bed design was modified, and a Thermal Enhancement Cartridge Heater Modified (TECH Mod) bed was developed. Internal electric cartridge heaters in the new design to improve end-of-life desorption, and also permit in-bed tritium accountability (IBA) calibration measurements to be made without the use of process tritium. Additional enhancements implemented into the TECH Mod design are also discussed. (authors)

Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

Energy Technology Data Exchange (ETDEWEB)

No, Hee Cheon; Park, Hyun Sik; Kim, Hyougn Tae; Moon, Young Min; Choi, Sung Won; Heo, Sun [Korea Advanced Institute Science and Technology, Taejon (Korea, Republic of)

1999-04-15

The loss-of-RHR accident during midloop operation has been important as results of the probabilistic safety analysis. The condensation models In RELAP5/MOD3 are not proper to analyze the midloop operation. To audit and improve the model in RELAP5/MOD3.2, several items of separate effect tests have been performed. The 29 sets of reflux condensation data is obtained and the correlation is developed with these heat transfer coefficient's data. In the experiment of the direct contact condensation in hot leg, the apparatus setting is finished and a few experimental data is obtained. Non-iterative model is used to predict the model in RELAP5/MOD3.2 with the results of reflux condensation and evaluates better than the present model. The results of the direct contact condensation in a hot leg represent to be similar with the present model. The study of the CCF and liquid entrainment in a surge line and pressurizer is selected as the third separate experiment and is on performance.

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

Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

Energy Technology Data Exchange (ETDEWEB)

Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-03-01

The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)

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 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)

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

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)

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

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

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

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

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

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

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

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

Thermal hydraulics of accelerator driven system: validation and analysis

International Nuclear Information System (INIS)

Kumari, I.; Khanna, A.

2014-01-01

This paper presents validation of RELAP5/Mod4.0 code modified to incorporate Lead Bismuth Eutectic (LBE)fluid properties for simulation of Accelerator Driven System (ADS) against Barone's NACIE facility.Results of mass flow rates (MFR), Reynolds number, heat transfer coefficients, temperatures and temperature difference for three powers (10.8, 21.7 and 32.5 kW) under natural circulation of LBE match with Barone's values within 7%,18%,37%, 5% and 8% of relative error respectively. After this validation Indian ADS for thermal power of 15 kW has been simulated. Simulated profiles of temperature, MFR and pressure drop LBE and air are reported. Air and LBE temperatures of present work match with literature design values within 5% of relative error. (author)

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.

Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

International Nuclear Information System (INIS)

Hwnag, M.

2001-04-01

The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicited with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applided for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is 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

Analysis of loss of offsite power transient using RELAP5/MOD1/NSC

International Nuclear Information System (INIS)

Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo

1986-01-01

System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAP5/MOD1/NSC), based upon the sequence of events for the KNU1( Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximate the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV setting etc. are recognized to be important in the transient analyses on a bestestimate basis. (Author)

LOFT transient thermal analysis for 10 inch primary coolant blowdown piping weld

International Nuclear Information System (INIS)

Howell, S.K.

1978-01-01

A flaw in a weld in the 10 inch primary coolant blowdown piping was discovered by LOFT personnel. As a result of this, a thermal analysis and fracture mechanics analysis was requested by LOFT personnel. The weld and pipe section were analyzed for a complete thermal cycle, heatup and Loss of Coolant Experiment (LOCE), using COUPLE/MOD2, a two-dimensional finite element heat conduction code. The finite element representation used in this analysis was generated by the Applied Mechanics Branch. The record of nodal temperatures for the entire transient was written on tape VSN=T9N054, and has been forwarded to the Applied Mechanics Branch for use in their mechanical analysis. Specific details and assumptions used in this analysis are found in appropriate sections of this report

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

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

Providing Home-Based HIV Testing and Counseling for Transgender Youth (Project Moxie): Protocol for a Pilot Randomized Controlled Trial.

Science.gov (United States)

Stephenson, Rob; Metheny, Nicholas; Sharma, Akshay; Sullivan, Stephen; Riley, Erin

2017-11-28

Transgender and gender nonconforming people experience some of the highest human immunodeficiency virus (HIV) rates in the United States, and experience many structural and behavioral barriers that may limit their engagement in HIV testing, prevention, and care. Evidence suggests that transgender and gender nonconforming youth (TY) are especially vulnerable to acquiring HIV, yet there is little research on TY and few services are targeted towards HIV testing, prevention, and care for this population. Telehealth presents an opportunity to mitigate some structural barriers that TY experience in accessing HIV testing, allowing TY to engage in HIV testing and counseling in a safe and nonjudgmental space of their choosing. Project Moxie is an HIV prevention intervention that pairs the use of HIV self-testing with remote video-based counseling and support from a trained, gender-affirming counselor. This study aims to offer a more positive HIV testing and counseling experience, with the goal of improving HIV testing frequency. Project Moxie involves a pilot randomized controlled trial (RCT) of 200 TY aged 15-24 years, who are randomized on a 1:1 basis to control or intervention arms. The aim is to examine whether the addition of counseling provided via telehealth, coupled with home-based HIV testing, can create gains in routine HIV testing among TY over a six-month follow-up period. This study implements a prospective pilot RCT of 200 TY recruited online. Participants in the control arm will receive one HIV self-testing kit and will be asked to report their results via the study's website. Participants in the experimental arm will receive one HIV self-testing kit and will test with a remotely-located counselor during a prescheduled video-counseling session. Participants are assessed at baseline, and at three and six months posttesting. Project Moxie was launched in June 2017 and recruitment is ongoing. As of August 21, 2017, the study had enrolled 130 eligible

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

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?...

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)

HANARO thermal hydraulic accident analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Chul; Kim, Heon Il; Lee, Bo Yook; Lee, Sang Yong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

For the safety assessment of HANARO, accident analyses for the anticipated operational transients, accident scenarios and limiting accident scenarios were conducted. To do this, the commercial nuclear reactor system code. RELAP5/MOD2 was modified to RELAP5/KMRR; the thermal hydraulic correlations and the heat exchanger model was changed to incorporate HANARO characteristics. This report summarizes the RELAP/KMRR calculation results and the subchannel analyses results based on the RELAP/KMRR results. During the calculation, major concern was placed on the integrity of the fuel. For all the scenarios, the important accident analysis parameters, i.e., fuel centerline temperatures and the minimum critical heat flux ratio(MCHFR), satisfied safe design limits. It was verified, therefore, that the HANARO was safely designed. 21 tabs., 89 figs., 39 refs. (Author) .new.

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

Innovation-driven efficient development of the Longwangmiao Fm large-scale sulfur gas reservoir in Moxi block, Sichuan Basin

Directory of Open Access Journals (Sweden)

Xinhua Ma

2016-03-01

Full Text Available The Lower Cambrian Longwangmiao Fm gas reservoir in Moxi block of the Anyue Gas field, Sichuan Basin, is the largest single-sandbody integrated carbonate gas reservoir proved so far in China. Notwithstanding this reservoir's advantages like large-scale reserves and high single-well productivity, there are multiple complicated factors restricting its efficient development, such as a median content of hydrogen sulfide, low porosity and strong heterogeneity of fracture–cave formation, various modes of gas–water occurrences, and close relation between overpressure and stress sensitivity. Up till now, since only a few Cambrian large-scale carbonate gas reservoirs have ever been developed in the world, there still exists some blind spots especially about its exploration and production rules. Besides, as for large-scale sulfur gas reservoirs, the exploration and construction is costly, and production test in the early evaluation stage is severely limited, all of which will bring about great challenges in productivity construction and high potential risks. In this regard, combining with Chinese strategic demand of strengthening clean energy supply security, the PetroChina Southwest Oil & Gas Field Company has carried out researches and field tests for the purpose of providing high-production wells, optimizing development design, rapidly constructing high-quality productivity and upgrading HSE security in the Longwangmiao Fm gas reservoir in Moxi block. Through the innovations of technology and management mode within 3 years, this gas reservoir has been built into a modern large-scale gas field with high quality, high efficiency and high benefit, and its annual capacity is now up to over 100 × 108 m3, with a desirable production capacity and development indexes gained as originally anticipated. It has become a new model of large-scale gas reservoirs with efficient development, providing a reference for other types of gas reservoirs in China.

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

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.

MOD-0A 200 kW wind turbine generator design and analysis report

Science.gov (United States)

Anderson, T. S.; Bodenschatz, C. A.; Eggers, A. G.; Hughes, P. S.; Lampe, R. F.; Lipner, M. H.; Schornhorst, J. R.

1980-01-01

The design, analysis, and initial performance of the MOD-OA 200 kW wind turbine generator at Clayton, NM is documented. The MOD-OA was designed and built to obtain operation and performance data and experience in utility environments. The project requirements, approach, system description, design requirements, design, analysis, system tests, installation, safety considerations, failure modes and effects analysis, data acquisition, and initial performance for the wind turbine are discussed. The design and analysis of the rotor, drive train, nacelle equipment, yaw drive mechanism and brake, tower, foundation, electricl system, and control systems are presented. The rotor includes the blades, hub, and pitch change mechanism. The drive train includes the low speed shaft, speed increaser, high speed shaft, and rotor brake. The electrical system includes the generator, switchgear, transformer, and utility connection. The control systems are the blade pitch, yaw, and generator control, and the safety system. Manual, automatic, and remote control are discussed. Systems analyses on dynamic loads and fatigue are presented.

Blowdown heat transfer surface in RELAP4/MOD6 and data comparisons

International Nuclear Information System (INIS)

Nelson, R.A.; Sullivan, L.H.

1978-01-01

RELAP4 is a thermal hydraulic analysis tool written to analyze transients in light water reactors (LWR). To date, most of the applications for RELAP4 have been to analyze postulated LOCA transients in LWR and the response of experimental systems to loss-of-coolant experiments. An important part of these analyses is the prediction of the fuel rod or heater surface temperature which involves the calculation of surface heat transfer coefficients. The paper describes the outcome of a significant blowdown heat transfer development effort which is incorporated in RELAP4/MOD6 (the current version of the code available to the United States public from the Argonne Code Center). The primary emphasis in the MOD6 development was on a PWR reflood capability. The best-estimate blowdown heat transfer correlation and logic were added to provide improved blowdown predictive capability

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

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.

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

Steady-state simulations of a 30-tube once-through steam generator with the RELAP5/MOD3 and RELAP5/MOD2 computer codes

International Nuclear Information System (INIS)

Hassan, Y.A.; Salim, P.

1991-01-01

This paper reports on a steady-state analysis of a 30-tube once-through steam generator that has been performed on the RELAPS/MOD3 and RELAPS/MOD2 computer codes for 100, 75, and 65% loads. Results obtained are compared with experimental data. The RELAP5/MOD3 results for the test facility generally agree reasonably well with the data for the primary-side temperature profiles. The secondary-side temperature profile predicted by RELAP5/MOD3 at 75 and 65% loads agrees fairly well with the data and is better than the RELAP5/MOD2 results. However, the RELAP5/MOD3 calculated secondary-side temperature profile does not compare well with the 100% load data

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

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

Energy Technology Data Exchange (ETDEWEB)

Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

International Nuclear Information System (INIS)

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-01-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds

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.

Development of the containment transient analysis code for the passive reactor

Energy Technology Data Exchange (ETDEWEB)

Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-05-01

This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.

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

Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

Directory of Open Access Journals (Sweden)

Muhammad Atta

2011-01-01

Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

Key seismic exploration technology for the Longwangmiao Fm gas reservoir in Gaoshiti–Moxi area, Sichuan Basin

Directory of Open Access Journals (Sweden)

Guangrong Zhang

2016-10-01

Full Text Available The dolomite reservoirs of the Lower Cambrian Longwangmiao Fm in the Gaoshiti–Moxi area, Sichuan Basin, are deeply buried (generally 4400–4900 m, with high heterogeneity, making reservoir prediction difficult. In this regard, key seismic exploration technologies were developed through researches. Firstly, through in-depth analysis on the existing geologic, drilling, seismic data and available research findings, basic surface and subsurface structures and geologic conditions within the study area were clarified. Secondly, digital seismic data acquisition technologies with wide azimuth, wide frequency band and minor bins were adopted to ensure even distribution of coverage of target formations through optimization of the 3D seismic geometry. In this way, high-accuracy 3D seismic data can be acquired through shallow, middle and deep formations. Thirdly, well-control seismic data processing technologies were applied to enhance the signal-to-noise ratio (SNR of seismic data for deep formations. Fourthly, a seismic response model was established specifically for the Longwangmiao Fm reservoir. Quantitative prediction of the reservoir was performed through pre-stack geo-statistics. In this way, plan distribution of reservoir thicknesses was mapped. Fifthly, core tests and logging data analysis were conducted to determine gas-sensitive elastic parameters, which were then used in pre-stack hydrocarbon detection to eliminate the multiple solutions in seismic data interpretation. It is concluded that application of the above-mentioned key technologies effectively promote the discovery of largescale marine carbonate gas reservoirs of the Longwangmiao Fm.

Development of regulatory technology for thermal-hydraulic safety analysis

International Nuclear Information System (INIS)

Bang, Young Seok; Lee, S. H.; Ryu, Y. H.

2001-02-01

The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

Energy Technology Data Exchange (ETDEWEB)

Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-04-15

The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.

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.

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

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

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)

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.

Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK

International Nuclear Information System (INIS)

Brus, N.A.; Ioussoupov, O.E.

2000-01-01

This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)

ATWS thermal-hydraulic analysis for Krsko Full Scope Simulator validation

International Nuclear Information System (INIS)

Parzer, I.; Kljenak, I.

2005-01-01

The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for annual ANSI/ANS validation of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD3.3 code and the input model for NPP Krsko, delivered by NPP Krsko, was used. In the presented paper the most severe ATWS scenario has been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied assuming AMSAC availability. (author)

Experiment data report for Semiscale Mod-1 Test S-05-3 (alternate ECC injection test)

International Nuclear Information System (INIS)

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

1977-03-01

Recorded test data are presented for Test S-05-3 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-05-3 was conducted from initial conditions of 2263 psia and 545 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 sides of the intact and broken loops and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. For Test S-05-3, specifically the effects of upper plenum coolant injection on core thermal and system response were being investigated

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

TC-13 Mod 0 and Mod 2 Steam Catapult Test Site

Data.gov (United States)

Federal Laboratory Consortium — Located on 11,000 feet of test runway, the TC-13 Mod 0 and Mod 2 Steam Catapult Test Site has in-ground catapults identical to those aboard carriers. This test site...

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.)

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)

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

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

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

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

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

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.

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)

Impact on the PSV Stuck Open according to the Henry-Fauske Model Modification in RELAP5/MOD3.3

Energy Technology Data Exchange (ETDEWEB)

Cho, Dong Hyun; Kim, Cheol Woo; Huh, Jae Young; Lee, Gyu Cheon; Kim, Shin Whan [KEPCO E-C, Daejeon (Korea, Republic of)

2014-10-15

Two different flow areas discharge same amount of design steam flow at the design condition but they provide the different flow rate during low pressure condition or two-phase mixture discharge. To evaluate the effect of the H-F model modification, the PSV stuck open event during a PSV popping test is selected since it involves the two phase discharge. In the present PSA practice for dealing with the variety of different plant operating states (POSs) during low power and shutdown (LPSD) operations, especially PSV popping test is performed during the POS2 of the overhaul period for OPR1000. To analyze thermal hydraulic behaviors of PSV stuck open event during POS2, RELAP5/MOD3.3 is used adopting the H-F critical flow model. In this paper, the impact on the PSV stuck open analysis during POS2 according to H-F critical flow model modification is investigated. Due to the modification of H-F model in RELAP5/MOD3.3 patch 4, the critical steam flow rate is increased at high pressure and thus the simulated PSV area is decreased. The change in PSV flow area impacts on the thermal hydraulic behaviors of the PSV stuck open event during POS2. PSA modeling can be changed depending on the results of thermal hydraulic analysis.

Impact on the PSV Stuck Open according to the Henry-Fauske Model Modification in RELAP5/MOD3.3

International Nuclear Information System (INIS)

Cho, Dong Hyun; Kim, Cheol Woo; Huh, Jae Young; Lee, Gyu Cheon; Kim, Shin Whan

2014-01-01

Two different flow areas discharge same amount of design steam flow at the design condition but they provide the different flow rate during low pressure condition or two-phase mixture discharge. To evaluate the effect of the H-F model modification, the PSV stuck open event during a PSV popping test is selected since it involves the two phase discharge. In the present PSA practice for dealing with the variety of different plant operating states (POSs) during low power and shutdown (LPSD) operations, especially PSV popping test is performed during the POS2 of the overhaul period for OPR1000. To analyze thermal hydraulic behaviors of PSV stuck open event during POS2, RELAP5/MOD3.3 is used adopting the H-F critical flow model. In this paper, the impact on the PSV stuck open analysis during POS2 according to H-F critical flow model modification is investigated. Due to the modification of H-F model in RELAP5/MOD3.3 patch 4, the critical steam flow rate is increased at high pressure and thus the simulated PSV area is decreased. The change in PSV flow area impacts on the thermal hydraulic behaviors of the PSV stuck open event during POS2. PSA modeling can be changed depending on the results of thermal hydraulic analysis

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...

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...

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

Scaling criteria and an assessment of Semiscale Mod-3 scaling for small-break loss-of-coolant transients

International Nuclear Information System (INIS)

Larson, T.K.; Anderson, J.L.; Shimeck, D.J.

1982-01-01

Various methods of scaling fluid thermal-hydraulic test facilities and their relative merits and disadvantages are examined in light of nuclear reactor safety considerations. Particular emphasis is placed on examination of the scaling of the Semiscale Mod-3 system and determination of thermal-hydraulic phenomena thought to be important during a small break loss-of-coolant accident in a pressurized water nuclear reactor. The influence of geometric and dynamic scaling concerns in the Mod-3 system on small break behavior are addressed from an engineering viewpoint and corrective measures contemplated or required to make results from Semiscale tests more meaningful relative to expected PWR response are discussed

Thermal-hydraulic calculation and analysis for QNPP (Qinshan Nuclear Power Plant) containment

International Nuclear Information System (INIS)

Xie Hui; Zhou Jie; He Yingchao

1993-01-01

Three containment thermal-hydraulic codes CONTEMPT-LT/028, CONTEMPT-4/MOD3 and COMPARE are used to compute and analyse the Qinshan Nuclear Power Plant (QNPP) containment response under LOCA or MSLB conditions. An evaluation of the capability of containment of QNPP is given

Thermal analysis of formation of nano-crystalline BaTiO3 using Ba(NO32 and TiO2

Directory of Open Access Journals (Sweden)

Md. Jawed Ansaree

2015-12-01

Full Text Available The reaction of Ba(NO32 with TiO2 was studied by thermogravimetric (TG and differential scanning calorimetric (DSC techniques up to 1000 °C and in nitrogen atmosphere. It was found that the formation of BaTiO3 takes place above 600 °C and that precursor mixing time and heating rate have no effect on the reaction temperature. BaTiO3 powder was prepared by calcination of Ba(NO32 and TiO2 precursor mixture at 800 °C for 8 h. X-ray diffraction analysis of the synthesized BaTiO3 confirmed the formation of tetragonal phase with lattice parameters a = 3.9950±0.0003 Å and c = 4.0318±0.0004 Å. Thermal analysis of the synthesized BaTiO3 powder showed weight loss within temperature range 40–1000 °C of only 0.40%. This small amount of weight loss was connected with some impurity phase, and identified as BaCO3 using Fourier transform infrared (FTIR technique.

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

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

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

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.

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.

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.

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

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.

Post-test simulation and analysis of the second full scale CHAN 28-element experiment (validations of CHAN-II (MOD 6) against experiments)

Energy Technology Data Exchange (ETDEWEB)

Bayoumi, M H; Muir, W C [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

An experimental program, the CHAN Thermal Chemical Experimental Program, has been setup at WNRE under COG/CANDEV to assess and verify the physical and mathematical models of the CHAN codes. The program has been progressing from studying separate effects in single-element experiments to a full integrated mode in a CANDU 28-element bundle geometry. The CHAN-II series codes are used in the licensing analysis of CANDU reactors. The basic code provides an efficient tool to predict the thermal response of a fuel channel during postulated loss-of-coolant accidents (LOCA) with and without a loss of emergency coolant injection (LOECI) in which the transport of heat by convection is greatly reduced. The code models the progression of the event including fuel channel geometry deformation due to severe overheating. It is the main objective of this paper to discuss further verification of the CHAN-II (MOD 6) computer code against the second full scale 28-element experiment performed at WNRE under COG/CANDEV, designed to represent a Pickering type bundle geometry. The main models and assumptions used in the code will be briefly described. The objective of the experiments is to provide data for the assessment of the physical and mathematical models of the CHAN codes and produce data for code verification under integrated conditions with significant hydrogen production and flow rates similar to the LOCA/LOECI scenario. The issue of whether the Zr/steam reaction is sustainable in a full bundle geometry at elevated temperatures is also examined. A comparison between the predictions of CHAN-II (MOD 6) and the experimental results is discussed. (author).12 refs., 17 figs.

Post-test simulation and analysis of the second full scale CHAN 28-element experiment (validations of CHAN-II (MOD 6) against experiments)

International Nuclear Information System (INIS)

Bayoumi, M.H.; Muir, W.C.

1995-01-01

An experimental program, the CHAN Thermal Chemical Experimental Program, has been setup at WNRE under COG/CANDEV to assess and verify the physical and mathematical models of the CHAN codes. The program has been progressing from studying separate effects in single-element experiments to a full integrated mode in a CANDU 28-element bundle geometry. The CHAN-II series codes are used in the licensing analysis of CANDU reactors. The basic code provides an efficient tool to predict the thermal response of a fuel channel during postulated loss-of-coolant accidents (LOCA) with and without a loss of emergency coolant injection (LOECI) in which the transport of heat by convection is greatly reduced. The code models the progression of the event including fuel channel geometry deformation due to severe overheating. It is the main objective of this paper to discuss further verification of the CHAN-II (MOD 6) computer code against the second full scale 28-element experiment performed at WNRE under COG/CANDEV, designed to represent a Pickering type bundle geometry. The main models and assumptions used in the code will be briefly described. The objective of the experiments is to provide data for the assessment of the physical and mathematical models of the CHAN codes and produce data for code verification under integrated conditions with significant hydrogen production and flow rates similar to the LOCA/LOECI scenario. The issue of whether the Zr/steam reaction is sustainable in a full bundle geometry at elevated temperatures is also examined. A comparison between the predictions of CHAN-II (MOD 6) and the experimental results is discussed. (author).12 refs., 17 figs

32 CFR 32.45 - Cost and price analysis.

Science.gov (United States)

2010-07-01

... 32 National Defense 1 2010-07-01 2010-07-01 false Cost and price analysis. 32.45 Section 32.45... price analysis. Some form of cost or price analysis shall be made and documented in the procurement files in connection with every procurement action. Price analysis may be accomplished in various ways...

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

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

Thermal characterization of QSH crashes in RFX-mod

Science.gov (United States)

Fassina, Alessandro; Gobbin, Marco; Franz, Paolo; Marrelli, Lionello; Ruzzon, Alberto

2012-10-01

QSH (Quasi Single Helicity) states have gained a growing interest in RFP research since they show improved confinement and transport features with respect to standard discharges. However, ITBs associated with QSH states can be obtained only in a transient way, and in general with a shorter lifetime with respect to that of the QSH phase [1]. In this work the analysis has essentially the purpose of confirming, with TS data, the Te dynamics seen with the double filter, multichord SXR spectrometer in [1]: TS data allow a better spatial definition of temperature profile and a more reliable description of plasma edge. Te profile features in rising and crashing phases are determined via ensemble averaging, possible precursors of thermal crashes are identified, while q(r) behavior is studied identifying the thermal structures associated with rational surfaces. [4pt] [1] Ruzzon et al, 39th EPS Conference, P2.023

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

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.

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

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

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

Analysis of the General Electric Company swell tests with RELAP4/MOD7

International Nuclear Information System (INIS)

Fischer, S.R.; Hendrix, C.E.

1979-01-01

The RELAP4/MOD7 nuclear reactor transient analysis code, presently being developed by EG and G Idaho, Inc., will incorporate several significant improvements over earlier versions of RELAP4. As part of the development of RELAP4/MOD7, a thorough assessment of the capability of the code to simulate water reactor LOCA phenomena is being made. This assessment is accomplished in part by comparing results from code calculations with test data from experimental facilities. Simulations of the General Electric Company (GE) level swell tests were performed as part of the code checkout. In these tests, a pressurized vessel partially filled with nearly saturated water was blown down through a simulated break located near the top of the vessel. Comparison of RELAP4 calculations with data from these experiments indicates that the code has the capability to model the unequal phase velocity flow and resulting density gradients that might occur in a BWR steam line break transient. Comparisons of RELAP4 calculations with data from two level swell experiments are presented

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.

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.

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.

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

FUDA MOD-2: a computer program for simulation the performance of fuel element validation exercise

International Nuclear Information System (INIS)

Chouhan, S.K.; Tripathi, R.M.; Prasad, P.N.; Chauhan, Ashok

2014-01-01

The PHWR fuel element performance is evaluated using the fuel analysis computer code FUDA MOD2. It is specifically written for performance simulation of UO 2 fuel pellet, located in zirconium alloy sheath operating under coolant pressure. For specific element power histories, the code investigates the variables and their interactions that govern fuel element performance. The input data requires pellet dimensions, element dimensions, sheath properties, heat transfer data, thermal hydraulic parameters of coolant, the inner filler gas composition, flux gradient and linear heat ratings (LHR) at different burn up. The output data generated by the code are radial temperature profile of fuel and sheath, fuel sheath-gap heat transfer coefficient, fission gas generated and released, fission gas pressure, sheath stress and strain for different burn-up zones. The code has been verified against literature data and post irradiation examinations carried out. It is also bench marked against various international fuel element simulation programmes available with water cooled reactors operating countries. The present paper describes the FUDA MOD2 code verification studies carried out using the literature data and post irradiation examination data. (author)

Experiment data report for Semiscale Mod-1 Test S-05-4 (alternate ECC injection test)

International Nuclear Information System (INIS)

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

1977-03-01

Recorded test data are presented for Test S-05-4 of the Semiscale Mod-1 alternate emergency core coolant 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-05-4 was conducted from initial conditions of 2266 psia and 543 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 each loop and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. The upper plenum coolant injection was scaled according to the heat stored in the metal mass of the upper plenum

TWODEE-2/MOD3, 2-D Time-Dependent Fuel Elements Thermal Analysis after PWR LOCA

International Nuclear Information System (INIS)

Lauben, G. N.

2001-01-01

1 - Description of problem or function: WREM-TOODEE2 is a two- dimensional, time-dependent, fuel-element thermal analysis program. Its primary purpose is to evaluate fuel-element thermal response during post-LOCA refill and reflood in a pressurized water reactor (PWR). 2 - Method of solution: TOODEE2 calculations are carried out in a two-dimensional mesh region defined in slab or cylindrical geometry by orthogonal grid lines. Coordinates which form order pairs are labeled x-y in slab geometry, and those in cylindrical geometry are labeled r-z for the axisymmetric case and r-theta for the polar case. Conduction and radiation are the only heat transfer mechanisms assumed within the boundaries of the mesh region. Convective and boiling heat transfer mechanisms are assumed at the boundaries. The program numerically solves the two-dimensional, time-dependent, heat conduction equation within the mesh region. 3 - Restrictions on the complexity of the problem: WREM-TOODEE2 considers only axisymmetric geometry although the equations for slab and polar geometry are included in the program

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.

RELAP5/MOD2 code assessment for the Semiscale Mod-2C Test S-LH-1

International Nuclear Information System (INIS)

Fineman, C.P.

1986-01-01

RELAP5/MOD2, Cycle 36.02, was assessed using data from Semiscale Mod-2C experiment S-LH-1. The major phenomena that occurred during the experiment were calculated by RELAP5/MOD2, although the duration and the magnitude of their effect on the transient were not always well calculated. Areas defined where further work was needed to improve the RELAP5 calculation include: (1) the system energy balance, (2) core interfacial drag, and 3) the heat transfer logic rod dryout criterion

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 of a best estimate auditing code for CANDU thermal hydraulic safety analysis

Energy Technology Data Exchange (ETDEWEB)

Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-04-01

The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)

Use of labelled atoms in thermal analysis

International Nuclear Information System (INIS)

Balek, V.; Beckman, I.N.

1985-01-01

The article informs of the preparation of labelled samples for which the most frequently used radionuclides are 14 C, 3 H or 2 H, 32 P, 35 S and others as well as radioactive gases such as 85 Kr, 133 Xe or 220 Rn and 222 Rn. The equipment is described for the application of labelled atoms in thermal analysis consisting of a detector for measuring radioactivity and a system for measuring thermal analysis parameters. Examples are given of the use of labelled atoms in the study of chemical reactions of solids, in autoradiography or in Moessbauer spectroscopy. The greatest attention is devoted to the use of labelled atoms in emanation thermal analysis. By this technique it is possible to study chemical reactions and phase transformations, to continuously monitor changes in the surface and morphology of dispersion substances, to characterize the mobility of defects in the structure of solids and the active state of the structure of solids and to ascertain mechanical, radiation and chemical effects on solids. Attention is also devoted to the technological applications of emanation thermal analysis (the solidification of cement paste, calcination and the firing of the mixture of oxides for the manufacture of ferrites). (E.S.)

The MOD-OA 200 kilowatt wind turbine generator design and analysis report

Science.gov (United States)

Andersen, T. S.; Bodenschatz, C. A.; Eggers, A. G.; Hughes, P. S.; Lampe, R. F.; Lipner, M. H.; Schornhorst, J. R.

1980-01-01

The project requirements, approach, system description, design requirements, design, analysis, system tests, installation safety considerations, failure modes and effects analysis, data acquisition, and initial performance for the MOD-OA 200 kw wind turbine generator are discussed. The components, the rotor, driven train, nacelle equipment, yaw drive mechanism and brake, tower, foundation, electrical system, and control systems are presented. The rotor includes the blades, hub and pitch change mechanism. The drive train includes the low speed shaft, speed increaser, high speed shaft, and rotor brake. The electrical system includes the generator, switchgear, transformer, and utility connection. The control systems are the blade pitch, yaw, and generator control, and the safety system. Manual, automatic, and remote control and Dynamic loads and fatigue are analyzed.

Experiment data report for Semiscale Mod-1 test S-28-3 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Gillins, R.L.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-3 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-28-3 was conducted from initial conditions of 15621 kPa and 555 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. Twelve steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

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

Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Esparza, V.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-4 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-28-4 was conducted from initial conditions of 15 646 kPa and 557 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. Thirty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

Energy Technology Data Exchange (ETDEWEB)

Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-04-01

The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

Analysis of Rotation and Transport Data in C-Mod ITB Plasmas

Science.gov (United States)

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

2009-11-01

Internal transport barriers (ITBs) spontaneously form near the half radius of Alcator C-Mod plasmas when the EDA H-mode is sustained for several energy confinement times in either off-axis ICRF heated discharges or in purely ohmic heated plasmas. These plasmas 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. It has long been observed that the intrinsic central plasma rotation that is strongly co-current following the H-mode transition slows and often reverses as the density peaks as the ITB forms. Recent spatial measurements demonstrate that the rotation profile develops a well in the core region that decreases continuously as central density rises while the value outside of the core remains strongly co-current. This results in the formation of a steep potential gradient/strong electric field at the location of the foot of the ITB density profile. The resulting E X B shearing rate is also quite significant at the foot. These analyses and the implications for plasma transport and stability will be presented.

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.

On mod 2 and higher elliptic genera

International Nuclear Information System (INIS)

Liu Kefeng

1992-01-01

In the first part of this paper, we construct mod 2 elliptic genera on manifolds of dimensions 8k+1, 8k+2 by mod 2 index formulas of Dirac operators. They are given by mod 2 modular forms or mod 2 automorphic functions. We also obtain an integral formula for the mod 2 index of the Dirac operator. As a by-product we find topological obstructions to group actions. In the second part, we construct higher elliptic genera and prove some of their rigidity properties under group actions. In the third part we write down characteristic series for all Witten genera by Jacobi theta-functions. The modular property and transformation formulas of elliptic genera then follow easily. We shall also prove that Krichever's genera, which come from integrable systems, can be written as indices of twisted Dirac operators for SU-manifolds. Some general discussions about elliptic genera are given. (orig.)

Kuosheng BWR/6 recirculation pump trip transient analysis with the RETRAN02/MOD5 code

International Nuclear Information System (INIS)

Wang, J.R.; Shih, C.

1992-01-01

A recirculation pump trip (RPT) event results in a reduction in recirculation flow, which reduces the core coolant flow rate. A reduction in core flow results in an increase in core void fraction and hence a decrease in core power due to negative void reactivity feedback. Although this category of events is less severe than others and generally considered as nonlimiting, core instability still may occur such as that at LaSalle on March 9, 1988. This paper focuses on the RPT transient analysis of Kuosheng Nuclear Power Plant (KNPP), which has two units of General Electric-designed boiling water reactor (BWR)/6 with rated core thermal power of 2894 MW and rated core flow of 10645 kg/s (23472 lb m /s). The approach to investigating the RPT transient of KNPP consists of two steps. The first step is to develop a plant-specific model using the RETRAN02/MOD5 code. In this step, various plant-specific information, including design documentation, drawings, safety analysis reports, and other information supplied by vendors were collected for model development. The RPT startup test at 68% power was used for system model benchmarking to ensure the adequacy of this model and identify several sensitive parameters. The second step is to assess whether similar power oscillation phenomena may occur at KNPP because of an RPT with isolated feedwater heater event. Two transient analyses (with or without reactor scram) of the KNPP RPT with isolated feedwater heater were investigated

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

Thermal-hydraulic studies on the safety of VVER-440 type nuclear power plants

International Nuclear Information System (INIS)

Tuunanen, J.

1994-01-01

The thesis includes several thermal-hydraulic analyses related to the Loviisa VVER-440 nuclear power plant units. The work consists of experimental studies, analysis of the experiments, analysis of some plant transients and development of a calculational model for calculation of boric concentrations in the reactor. In the first part of thesis, in the case of simulation of boric acid solution behaviour during long-term cooling period of LOCAs, experiments were performed in scaled-down test facilities. The experimental data together with the results of RELAP5/MOD3 simulations were used to develop a model for calculations of boric acid concentrations in the reactor during LOCAs. In the second part, in the case of simulation of horizontal generators, experiments were performed with PACTEL integral test loop to simulate loss of feedwater transients. The PACTEL experiments as well as earlier REWETT-III natural circulation tests, were analyzed with RELAP5/MOD3 Version 5m5 code. The third part of the work consists of simulations of Loviisa VVER reactor pump trip transients with RELAP5/MOD1-Eur, RELAP5/MOD3 and CATHARE codes. (56 refs., 9 figs.)

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.

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.

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

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

Experiment data report for Semiscale Mod-1 Test S-28-1 (steam generator tube rupture test series)

International Nuclear Information System (INIS)

Collins, B.L.; Coppin, C.E.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-1 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-28-1 was conducted from initial conditions of 15 767 kPa and 557 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. Sixty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

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

Timing analysis of PWR fuel pin failures

International Nuclear Information System (INIS)

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

1992-09-01

Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report

RELAP5/Mod3.3 and MARS3.0a Modeling of a Siphon Break Experiment

International Nuclear Information System (INIS)

Park, Su Ki; Kim, Heon Il; Park, Cheol; Yoon, Ju Hyeon

2011-01-01

Pool water plays a very important role as a final heat sink for most pool-type research reactors following postulated events. Therefore, one of design criteria for the reactors is that the water level of reactor pool must not decrease below a predefined elevation even against the most severe accident due to ruptures of coolant boundary of connecting systems to the reactor pool. In order to accomplish the design criterion, all the connecting systems are usually arranged to be above the elevation of reactor core. However, some research reactors with a downward flow in the reactor core have a primary cooling system located below the elevation of reactor core because of meeting an available net positive suction head of pumps in the system. These reactors have a provision consisting of pipes penetrating a reactor pool wall at a higher elevation than that of reactor core and siphon break devices to meet the design criterion. A series of experiments was carried out to figure out thermal hydraulic characteristics during siphon is blocked and establish design requirements for siphon breaker. The experimental study provided a lot of data and observations to the process of siphon break, but it does not provide a sufficient theoretical analysis and present practical design requirements applicable to industry. The experimental range is not also sufficient to cover operating conditions of siphon breakers for research reactors. A series of numerical simulations on the experimental data has been tried by using thermal hydraulic system analysis codes, RELAP5/Mod3.3 and MARS3.0a. This paper includes a part of the numerical simulations. First output from this study shows an importance of an adequate use of thermal hydraulic models in the codes and a big different prediction between the two codes especially in relation to the use of choked flow option. From this study, it seems that RELAP5/Mod3.3 has some problems on the control of a choked flow option-flag or the prediction of a

Peach bottom instability analysis with a RELAP5/PARCSv2.7 detailed thermal-hydraulic–neutronic model

International Nuclear Information System (INIS)

Abarca, A.; Barrachina, T.; Miró, R.; Ginestar, D.; Verdú, G.

2012-01-01

Highlights: ► A RELAP5-MOD3.3/PARCSv2.7 model developed to characterize oscillations in BWR. ► The TH to neutronic mapping is based on the Lambda modes obtained with VALKIN code. ► The results show that point PT U PV is an unstable point with a bottom-peaked profile. ► The SSA and the Power Modal Decomposition have been applied to the LPRMs data. ► An in-phase coupled with an out-of-phase oscillation appears. - Abstract: In this work, BWR stability analysis has been performed on an operating point (PT U PV) of Peach Bottom NPP which is inside the exclusion region in the operating power-flow map. The simulation has been made with the coupled code RELAP5-MOD3.3/PARCSv2.7. This point is achieved departing from test point 3 by a control rod movement as it is usually performed in Nuclear Power Plants. The thermal-hydraulic model is a detailed model that includes all the reactor vessel components: jet pumps, recirculation pumps, downcomer, reactor core and also the separator and the dryer. The reactor core has been modeled with 72 thermal-hydraulic channels, 71 represent the active core and 1 represents the core bypass. The reactor core thermal-hydraulic to neutronic representation (mapping) has been divided in four quadrants according to the first and second power harmonics (Lambda modes) obtained previously with the VALKIN code. This mapping was chosen in order not to condition the oscillation pattern. The transient starts with the control rod movement. The calculated results show that point PT U PV is an unstable point and the obtained relative axial power distribution shows a bottom-peaked profile, which is characteristic of unstable cores.

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)

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

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

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

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)

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

International Nuclear Information System (INIS)

Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

Control of internal transport barriers on Alcator C-Mod

International Nuclear Information System (INIS)

Fiore, C.L.; Bonoli, P.T.; Ernst, D.R.; Hubbard, A.E.; Greenwald, M.J.; Lynn, A.; Marmar, E.S.; Phillips, P.; Redi, M.H.; Rice, J.E.; Wolfe, S.M.; Wukitch, S.J.; Zhurovich, K.

2004-01-01

Recent studies of internal transport and double transport barrier regimes in the Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] have explored the limits for forming, maintaining, and controlling these plasmas. The C-Mod provides a unique platform for studying such discharges: the ions and electrons are tightly coupled by collisions and the plasma has no internal particle or momentum sources. The double-barrier mode comprised of an edge barrier with an internal transport barrier (ITB) can be induced at will using off-axis ion cyclotron range of frequency (ICRF) injection on either the low or high field side of the plasma with either of the available ICRF frequencies (70 or 80 MHz). When an enhanced D α high confinement mode (EDA H-mode) is accessed in Ohmic plasmas, the double barrier ITB forms spontaneously if the H-mode is sustained for ∼2 energy confinement times. The ITBs formed in both Ohmic and ICRF heated plasmas are quite similar regardless of the trigger method. They are characterized by strong central peaking of the electron density, and a reduction of the core particle and energy transport. The control of impurity influx and heating of the core plasma in the presence of the ITB have been achieved with the addition of central ICRF power in both the Ohmic H-mode and ICRF induced ITBs. The radial location of the particle transport barrier is dependent on the toroidal magnetic field but not on the location of the ICRF resonance. A narrow region of decreased electron thermal transport, as determined by sawtooth heat pulse analysis, is found in these plasmas as well. Transport analysis indicates that a reduction of the particle diffusivity in the barrier region allows the neoclassical pinch to drive the density and impurity accumulation in the plasma center. An examination of the gyrokinetic stability at the trigger time for the ITB suggests that the density and temperature profiles are inherently stable to ion temperature gradient and

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

Mod-1 wind turbine generator analysis and design report, volume 1

Science.gov (United States)

1979-01-01

The activities leading to the completion of detail design of the MOD-1 wind turbine generator are described. Emphasis is placed on the description of the design as it finally evolved. However, the steps through which the design progressed are also traced in order to understand the major design decisions.

Thermal behavior of the Medicina 32-meter radio telescope

Science.gov (United States)

Pisanu, Tonino; Buffa, Franco; Morsiani, Marco; Pernechele, Claudio; Poppi, Sergio

2010-07-01

We studied the thermal effects on the 32 m diameter radio-telescope managed by the Institute of Radio Astronomy (IRA), Medicina, Bologna, Italy. The preliminary results show that thermal gradients deteriorate the pointing performance of the antenna. Data has been collected by using: a) two inclinometers mounted near the elevation bearing and on the central part of the alidade structure; b) a non contact laser alignment optical system capable of measuring the secondary mirror position; c) twenty thermal sensors mounted on the alidade trusses. Two series of measurements were made, the first series was performed by placing the antenna in stow position, the second series was performed while tracking a circumpolar astronomical source. When the antenna was in stow position we observed a strong correlation between the inclinometer measurements and the differential temperature. The latter was measured with the sensors located on the South and North sides of the alidade, thus indicating that the inclinometers track well the thermal deformation of the alidade. When the antenna pointed at the source we measured: pointing errors, the inclination of the alidade, the temperature of the alidade components and the subreflector position. The pointing errors measured on-source were 15-20 arcsec greater than those measured with the inclinometer.

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.

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.

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.

Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty

International Nuclear Information System (INIS)

Borges, Ronaldo Celem

2001-10-01

This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

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

An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

International Nuclear Information System (INIS)

Hall, D.G.; Watkins, J.C.

1987-01-01

This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

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

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

Assessment of BETHSY Test 9.1.b using RELAP5/MOD3

International Nuclear Information System (INIS)

Lee, S.; Chung, B.D.; Kim, H.J.

1993-06-01

The 2'' cold leg break test 9.l.b, conducted at the BETHSY facility was analyzed using the RELAP5/MOD3 Version 5m5 code. The test 9.l.b was conducted with the main objective being the investigation of the thermal-hydraulic mechanisms responsible for the large core uncovery and fuel heat-up, requiring the implementation of an ultimate procedure. The present analysis demonstrates the code's capability to predict, with sufficient accuracy, the main phenomena occurring in the depressurization transient, both from a qualitative and quantitative point of view. Nevertheless, several differences regarding the evolution of phenomena and affecting the timing order have to be pointed out in the base calculation. Three calculations were carried out to study the sensitivity to change of the nodalization in the components of the loop seal cross-over legs, and of the auxiliary feedwater control logics, and of the break discharge coefficient

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

Photocatalytic Removal of Dye and Reaction Mechanism Analysis over Y2O3 Composite Nanomaterials

Directory of Open Access Journals (Sweden)

Guo Xiaohui

2017-01-01

Full Text Available Y2O3 supported photocatalysts MOx (M = Fe, Ti and Bi/Y2O3 were synthesized by hydrothermal and deposition-precipitation method and used in photocatalytic degradation of xylenol orange and rhodamine B under UV light irradiation. The crystalline structure and optical properties were well characterized by XRD and UV-vis DRS. The results of XRD revealed that the MOx/Y2O3 were composed of Fe2O3, TiO2, Bi0.8Y1.2O3 and Y2O3. The UV-vis DRS showed that MOx/Y2O3 photocatalysts exhibited stronger absorption in ultraviolet, and the absorption edge shifted to visible light region significantly. The photocatalytic experiments indicated that MOx/Y2O3 photocatalysts showed better activity for photodegradation of xylenol orange than rhodamine B. Moreover, the role of Y2O3 and the mechanism of photocatalysis is proposed.

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...

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.

Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

Energy Technology Data Exchange (ETDEWEB)

Perneczky, L. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

1997-12-31

Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

Energy Technology Data Exchange (ETDEWEB)

Perneczky, L [KFKI Atomic Energy Research Inst., Budabest (Hungary)

1998-12-31

Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

Preparation and comparitive analysis of MCNP thermal libraries for liquid hydrogen and deuterium using NJOY97 on 32 bit and 64 bit computers

International Nuclear Information System (INIS)

Jo, Y. S.; Kim, J. D.; Kil, C. S.; Jang, J. H.

1999-01-01

The scattering laws and MCNP thermal libraries for liquid hydrogen and deuterium are comparatively calculated on HP715 (32-bit computer) and SGI IP27 (64-bit computer) using NJOY97. The results are also compared with the experimental data. In addition, MCNP calculations for the nuclear design of a cold neutron source at HANARO are performed with the newly generated MCNP thermal libraries from two different computers and the results are compared

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

Thermal Properties and Thermal Analysis:

Science.gov (United States)

Kasap, Safa; Tonchev, Dan

The chapter provides a summary of the fundamental concepts that are needed to understand the heat capacity C P, thermal conductivity κ, and thermal expansion coefficient α L of materials. The C P, κ, and α of various classes of materials, namely, semiconductors, polymers, and glasses, are reviewed, and various typical characteristics are summarized. A key concept in crystalline solids is the Debye theory of the heat capacity, which has been widely used for many decades for calculating the C P of crystals. The thermal properties are interrelated through Grüneisen's theorem. Various useful empirical rules for calculating C P and κ have been used, some of which are summarized. Conventional differential scanning calorimetry (DSC) is a powerful and convenient thermal analysis technique that allows various important physical and chemical transformations, such as the glass transition, crystallization, oxidation, melting etc. to be studied. DSC can also be used to obtain information on the kinetics of the transformations, and some of these thermal analysis techniques are summarized. Temperature-modulated DSC, TMDSC, is a relatively recent innovation in which the sample temperature is ramped slowly and, at the same time, sinusoidally modulated. TMDSC has a number of distinct advantages compared with the conventional DSC since it measures the complex heat capacity. For example, the glass-transition temperature T g measured by TMDSC has almost no dependence on the thermal history, and corresponds to an almost step life change in C P. The new Tzero DSC has an additional thermocouple to calibrate better for thermal lags inherent in the DSC measurement, and allows more accurate thermal analysis.

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

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.

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.

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.

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.

Internal transport barrier production and control in Alcator C-Mod

Energy Technology Data Exchange (ETDEWEB)

Fiore, C L [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Bonoli, P T [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Ernst, D R [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Greenwald, M J [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Marmar, E S [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Redi, M H [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Rice, J E [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Wukitch, S J [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Zhurovich, K [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

2004-12-01

Internal transport barriers (ITBs), marked by a steep density profile, even stronger peaking in the pressure profile and reduction of core transport are obtained in Alcator C-Mod. They are induced by the use of off-axis D(H) ICRF (ion cyclotron range of frequencies) power deposition. They also arise spontaneously in Ohmic H-mode plasmas once the H-mode lasts for several energy confinement times. Recent studies have explored the limits for forming, maintaining and controlling these plasmas. The C-Mod provides a unique platform for studying such discharges: the high density (up to 8 x 10{sup 20} m{sup -3}) causes the ions and electrons to be tightly coupled by collisions with T{sub i}/T{sub e} = 1, and the plasma has no internal particle or momentum sources. The ITBs formed in both Ohmic and ICRF heated plasmas are quite similar regardless of the trigger method. Control of impurity influx and heating of the core plasma in the presence of the ITB have been achieved with the addition of central ICRF power, in both Ohmic H-mode and ICRF induced ITBs. Control of the radial location of the transport barrier is achieved through manipulation of the toroidal magnetic field and plasma current. A narrow region of decreased electron thermal transport, as determined by sawtooth heat pulse analysis, is found in these plasmas as well. Transport analysis indicates that reduction of the particle diffusivity in the barrier region allows the neoclassical pinch to drive the density and impurity accumulation in the plasma centre. Examination of the gyro-kinetic stability indicates that the density and temperature profiles of the plasma core are inherently stable to long-wavelength drift mode driven turbulence at the onset time of the ITB, but that the increasing density gradients cause the trapped electron mode to play a role in providing a control mechanism to ultimately limit the density and impurity rise in the plasma centre.

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)

Analysis of KNU1 loss of normal feedwater

International Nuclear Information System (INIS)

Kim, Hho-Jung; Chung, Bub-Dong; Lee, Young-Jin; Kim, Jin-Soo

1986-01-01

Simulation of the system thermal-hydraulic parameters was carried out following the KNU1 (Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on November 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS (Reactor Coolant system) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018. (author)

NetMOD Version 2.0 Parameters

Energy Technology Data Exchange (ETDEWEB)

Merchant, Bion J. [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 probability of detection can be computed given a detection threshold. This document describes the parameters that are used to configure the NetMOD tool and the input and output parameters that make up the simulation definitions.

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

Application of RELAP/SCDAPSIM with integrated uncertainty options to research reactor systems thermal hydraulic analysis

International Nuclear Information System (INIS)

Allison, C.M.; Hohorst, J.K.; Perez, M.; Reventos, F.

2010-01-01

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 the international SCDAP 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 publicly available RELAP5 and SCDAP models in combination with advanced programming and numerical techniques and other SDTP-member modeling/user options. One such member developed option is an integrated uncertainty analysis package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). This paper briefly summarizes the features of RELAP/SCDAPSIM/MOD4.0 and the integrated uncertainty analysis package, and then presents an example of how the integrated uncertainty package can be setup and used for a simple pipe flow problem. (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.

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

Effects of maturity at harvest on the nutritive value and ruminal digestion of Eragrostis tef (cv. Moxie) when fed to beef cattle.

Science.gov (United States)

Vinyard, J R; Hall, J B; Sprinkle, J E; Chibisa, G E

2018-05-18

Teff (Eragrostis tef cv. Moxie), a warm-season annual grass, could be an excellent forage for beef cattle. However, there is limited information on its nutritive value to cattle when harvested at different stages of maturity. Thus, the objective of this research was to determine the effect of feeding teff hay harvested at the boot (BT), early-heading (EH), or late-heading (LH) stages of maturity on nutrient intake, ruminal fermentation characteristics, omasal nutrient flow, and N utilization in beef cattle. Six ruminally cannulated beef heifers (mean initial BW ± SD, 476 ± 32.6) were used in a replicated 3 × 3 Latin square design with 28-d periods (18 d for adaptation and 10 d for measurements). Dry matter intake was measured daily. Indwelling pH loggers were used to measure ruminal pH from d 21 to 28. Ruminal fluid and omasal digesta were collected from d 26 to 28 to determine fermentation characteristics and omasal nutrient flow. Fecal and urine samples to quantify N excretion were also collected (d 26 to 28). Blood samples for plasma urea-N (PUN) determination were collected 3 h post-feeding on d 28. There were no changes (P > 0.28) in the ADF or NDF content of teff with advancing maturity, but iNDF increased (P ruminal total short-chain fatty acid (SCFA) concentration, pH, digestibility and outflow of DM, OM, NDF, ADF, and CP. However, the CP content of BT hay was greater (P ruminal ammonia-N (NH3-N) concentration was greater (P ruminal NH3-N concentration, feeding beef heifers EH and LH compared to BT teff hay did not compromise ruminal digestion and outflow of DM, OM, NDF, ADF, and CP, and microbial protein synthesis. Advancing maturity in teff hay also resulted in a decrease in the excretion of total N and urine N and UUN when fed to cattle.

Implementation of a nonequilibrium condensation model in RELAP4/MOD7

International Nuclear Information System (INIS)

Fischer, S.R.; Chow, H.; Van Arsdall, G.

1979-01-01

RALAp, which is used to simulate the thermal hydraulic behavior of light water reactors subjected to various LOCA transients, is based on the assumption of thermodynamic equilibrium between liquid and vapor within fluid volumes. This assumption, while being appropriate for much of a LOCA transient, is not adequate during the ECC accumulator injection phase as determined by comparisons of code calculations with experimental data. To overcome this limitation, a general model to simulate the nonequilibrium phenomena associated with the mixing of subcooled water with saturated steam has been developed and is operational on preliminary versions of RELAP4/MOD7

Mechanisms for ITB formation and control in Alcator C-Mod identified through gyrokinetic simulations of TEM turbulence

International Nuclear Information System (INIS)

Ernst, D.R.; Basse, N.; Bonoli, P.T.; Catto, P.J.; Fiore, C.L.; Greenwald, M.; Hubbard, A.E.; Marmar, E.S.; Porkolab, M.; Rice, J.E.; Zeller, K.; Zhurovich, K.; Dorland, W.

2005-01-01

Internal particle and thermal energy transport barriers are produced in Alcator C-Mod with off-axis ICRF heating, with core densities exceeding 10 21 m -3 , without core fueling, and with little change in the temperature profile. Applying on-axis ICRF heating controls the core density gradient and rate of rise. The present study employs linear and nonlinear gyrokinetic simulations of trapped electron mode (TEM) turbulence to explore mechanisms for ITB formation and control in Alcator C-Mod ITB experiments. Anomalous pinches are found to be negligible in our simulations; further, the collisional Ware pinch is sufficient to account for the slow density rise, lasting many energy confinement times. The simulations have revealed new nonlinear physics of TEM turbulence. The critical density gradient for onset of TEM turbulent transport is nonlinearly up-shifted by zonal flows. As the density profile peaks, during ITB formation, this nonlinear critical gradient is eventually exceeded, and the turbulent particle diffusivity from GS2 gyrokinetic simulations matches the particle diffusivity from transport analysis, within experimental errors. A stable equilibrium is then established when the TEM turbulent diffusion balances the Ware pinch in the ITB. This equilibrium is sensitive to temperature through gyroBohm scaling of the TEM turbulent transport, and the collisionality dependence of the neoclassical pinch, providing for control of the density rate of rise with on-axis RF heating. With no core particle fueling, and ∼1 mm between density spatial channels, the C-Mod experiments provide a nearly ideal test bed for particle transport studies. The pure TEM is the only unstable drift mode in the ITB, producing particle transport driven by the density gradient. (author)

CuI-catalyzed photochemical or thermal reactions of 3-(2-azidobenzylidene)lactams. Application to the synthesis of fused indoles.

Science.gov (United States)

Shi, Zongjun; Ren, Yuwei; Li, Bing; Lu, Shenci; Zhang, Wei

2010-06-14

Photochemical or thermal reactions of 3-(2-azidobenzylidene)-lactams afforded fused indoles such as indolo[3,2-c]quinolin-6-ones, pyrido[4,3-b]indol-1-ones and other similar compounds in moderate to high yields via cyclization-ring expansion reactions. The photolytic process was much more facile than the thermal process and could be further improved by addition of CuI.

RELAP4/MOD6 reflood heat transfer and data comparison

International Nuclear Information System (INIS)

Nelson, R.A.; Sullivan, L.H.

1981-01-01

This discussion of RELAP4/MOD6 will be limited to the reflood heat transfer models and evaluation of these models by comparison of calculation with results from three reflood experiments. The discussion of the model includes the heat transfer surface concept, the heat transfer correlations, the superheat model and the entrainment model which presents both the two-phase heat transfer and hydraulic models. In the discussion of the reflood heat transfer, the mathematical concept of a multidimensional surface is used to represent the heat flux of a given heat transfer correlation or correlations dependent upon such variables as quality, wall superheat and flux. This concept has been used to investigate the characteristics of the correlations, which are discusssed in detail, and the way they are applied to the two-phase mixture. Of primary importance in the reflood core heat transfer is the consideration of thermal nonequilibrium between the phases and the liquid entrainment, and its distribution up the core. Results obtained to date show the heat transfer and hydraulics to be closely coupled. Comparison of the RELAP4/MOD6 reflood calculations with the data from the forced feed FLECHT and gravity feed FLECHT-SET and Semiscale reflood experiments indicates that the heat transfer and hydraulic models are operational and yield good results

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

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.

VARSKIN MOD 2 and SADDE MOD2: Computer codes for assessing skin dose from skin contamination

International Nuclear Information System (INIS)

Durham, J.S.

1992-12-01

The computer code VARSKIN has been modified to calculate dose to skin from three-dimensional sources, sources separated from the skin by layers of protective clothing, and gamma dose from certain radionuclides correction for backscatter has also been incorporated for certain geometries. This document describes the new code, VARSKIN Mod 2, including installation and operation instructions, provides detailed descriptions of the models used, and suggests methods for avoiding misuse of the code. The input data file for VARSKIN Mod 2 has been modified to reflect current physical data, to include the contribution to dose from internal conversion and Auger electrons, and to reflect a correction for low-energy electrons. In addition, the computer code SADDE: Scaled Absorbed Dose Distribution Evaluator has been modified to allow the generation of scaled absorbed dose distributions for mixtures of radionuclides and intereat conversion and Auger electrons. This new code, SADDE Mod 2, is also described in this document. Instructions for installation and operation of the code and detailed descriptions of the models used in the code are provided

Mod-5A wind turbine generator program design report. Volume 3: Final design and system description, 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. Volume 3, book 1 describes the performance and characteristics of the MOD-5A wind turbine generator in its final configuration. Each subsystem - the rotor, drivetrain, nacelle, tower and foundation is described in detail.

Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

2003-01-01

The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

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

Seismic prediction on the favorable efficient development areas of the Longwangmiao Fm gas reservoir in the Gaoshiti–Moxi area, Sichuan Basin

Directory of Open Access Journals (Sweden)

Guangrong Zhang

2017-05-01

Full Text Available The Lower Cambrian Longwangmiao Fm gas reservoir in the Gaoshiti–Moxi area, the Sichuan Basin, is a super giant monoblock marine carbonate gas reservoir with its single size being the largest in China. The key to the realization of high and stable production gas wells in this gas reservoir is to identify accurately high-permeability zones where there are dissolved pores or dissolved pores are superimposed with fractures. However, high quality dolomite reservoirs are characterized by large burial depth and strong heterogeneity, so reservoir prediction is of difficult. In this paper, related seismic researches were carried out and supporting technologies were developed as follows. First, a geologic model was built after an analysis of the existing data and forward modeling was carried out to establish a reservoir seismic response model. Second, by virtue of well-oriented amplitude processing technology, spherical diffusion compensation factor was obtained based on VSP well logging data and the true amplitude of seismic data was recovered. Third, the resolution of deep seismic data was improved by using the well-oriented high-resolution frequency-expanding technology and prestack time migration data of high quality was acquired. And fourth, multiple shoal facies reservoirs were traced by using the global automatic seismic interpretation technology which is based on stratigraphic model, multiple reservoirs which are laterally continuous and vertically superimposed could be predicted, and the areal distribution of high quality reservoirs could be described accurately and efficiently. By virtue of the supporting technologies, drilling trajectory is positioned accurately, and the deployed development wells all have high yield. These technologies also promote the construction of a modern supergiant gas field of tens of billions of cubic meters.

Study on application of two-fluid model in narrow annular channel

International Nuclear Information System (INIS)

Chen Jun; Yang Yanhua; Zhao Hua

2007-01-01

The Chexal-Harrison two-phase wall and inter-phase friction models developed by EPRI newly and the simple two-phase wall and inter-phase heat transfer models put forward by the paper are used to set up the two-fluid model which is fitted for boiling heat transfer and flow in narrow annular channel. On the base of the two-fluid model, a thermal hydraulic code-THYME is accomplished. Then the thermal hydraulic characteristic of narrow annular channel is analyzed by RELAP5/MOD3.2 code and THYME code. Compared with experimental data, RELAP5/MOD3.2 underestimates the outlet steam, and the results of THYME is agreed with the experimental data. (authors)

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).

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.

Countercurrent flow limitation model for RELAP5/MOD3

International Nuclear Information System (INIS)

Riemke, R.A.

1991-01-01

This paper reports on a countercurrent flow limitation model incorporated into the RELAP5/MOD3 system transient analysis code. The model is implemented in a manner similar to the RELAP5 chocking model. Simulations using air/water flooding test problem demonstrate the ability of the code to significantly improve its comparison to data when a flooding correlation is used

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

Hot particle dose calculations using the computer code VARSKIN Mod 2

International Nuclear Information System (INIS)

Durham, J.S.

1991-01-01

The only calculational model recognised by the Nuclear Regulatory Commission (NRC) for hot particle dosimetry is VARSKIN Mod 1. Because the code was designed to calculate skin dose from distributed skin contamination and not hot particles, it is assumed that the particle has no thickness and, therefore, that no self-absorption occurs within the source material. For low energy beta particles such as those emitted from 60 Co, a significant amount of self-shielding occurs in hot particles and VARSKIN Mod 1 overestimates the skin dose. In addition, the presence of protective clothing, which will reduce the calculated skin dose for both high and low energy beta emitters, is not modelled in VARSKIN Mod 1. Finally, there is no provision in VARSKIN Mod 1 to calculate the gamma contribution to skin dose from radionuclides that emit both beta and gamma radiation. The computer code VARSKIN Mod 1 has been modified to model three-dimensional sources, insertion of layers of protective clothing between the source and skin, and gamma dose from appropriate radionuclides. The new code, VARSKIN Mod 2, is described and the sensitivity of the calculated dose to source geometry, diameter, thickness, density, and protective clothing thickness are discussed. Finally, doses calculated using VARSKIN Mod 2 are compared to doses measured from hot particles found in nuclear power plants. (author)

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...

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.

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

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

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

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

Design of a New Optical System for Alcator C-Mod Motional Stark Effect Diagnostic

International Nuclear Information System (INIS)

Ko, Jinseok; Scott, Steve; Bitter, Manfred; Lerner, Scott

2009-01-01

The motional Stark effect (MSE) diagnostic on Alcator C-Mod uses an in-vessel optical system (five lenses and three mirrors) to relay polarized light to an external polarimeter because port access limitations on Alcator C-Mod preclude a direct view of the diagnostic beam. The system experiences unacceptable, spurious drifts of order several degrees in measured pitch angle over the course of a run day. Recent experiments illuminated the MSE diagnostic with polarized light of fixed orientation as heat was applied to various optical elements. A large change in measured angle was observed as two particular lenses were heated, indicating that thermal-stress-induced birefringence is a likely cause of the spurious variability. Several new optical designs have been evaluated to eliminate the affected in-vessel lenses and to replace the focusing they provide with curved mirrors; however, ray tracing calculations imply that this method is not feasible. A new approach is under consideration that utilizes in situ calibrations with in-vessel reference polarized light sources. 2008 American Institute of Physics.

Development of a nuclear power plant system analysis code

International Nuclear Information System (INIS)

Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

1997-07-01

During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

Studies of thermal stratification in water pool

International Nuclear Information System (INIS)

Verma, P.K.; Chandraker, D.K.; Nayak, A.K.; Vijayan, P.K.

2015-01-01

Large water pools are used as a heat sink for various cooling systems used in industry. In context of advance nuclear reactors like AHWR, it is used as ultimate heat sink for passive systems for decay heat removal and containment cooling. This system incorporates heat exchangers submerged in the large water pool. However, heat transfer by natural convection in pool poses a problem of thermal stratification. Due to thermal stratification hot layers of water accumulate over the relatively cold one. The heat transfer performance of heat exchanger gets deteriorated as a hot fluid envelops it. In the nuclear reactors, the walls of the pool are made of concrete and it may subject to high temperature due to thermal stratification which is not desirable. In this paper, a concept of employing shrouds around the heat source is studied. These shrouds provide a bulk flow in the water pool, thereby facilitating mixing of hot and cold fluid, which eliminate stratification. The concept has been applied to the a scaled model of Gravity Driven Water Pool (GDWP) of AHWR in which Isolation Condensers (IC) tubes are submerged for decay heat removal of AHWR using ICS and thermal stratification phenomenon was predicted with and without shrouds. To demonstrate the adequacy of the effectiveness of shroud arrangement and to validate the simulation methodology of RELAP5/Mod3.2, experiments has been conducted on a scaled model of the pool with and without shroud. (author)

Numerical modelling of ICRF physics experiments in the Alcator C-mod tokamak

International Nuclear Information System (INIS)

Bonoli, P.T.; Boivin, R.L.; Brambilla, M.

2001-01-01

A full-wave spectral code (TORIC) has been used to simulate mode converted ion Bernstein wave (IBW) propagation and absorption for the first time at high poloidal mode number (-80< m<+80). Converged wave solutions for the mode converted wave are obtained in this limit and the predicted electron damping of the IBW is found to be consistent with experimental measurements from the Alcator C-Mod tokamak. The TORIC code has also been coupled to a bounce-averaged Fokker Planck module FPPRF and the combined codes are now run within the transport analysis tool TRANSP. This model was used to analyze off-axis hydrogen minority heating experiments in C-Mod where an internal transport barrier was obtained. (author)

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...

Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

Energy Technology Data Exchange (ETDEWEB)

No, Hee Cheon; Moon, Young Min; Lee, Dong Won; Lee, Sang Ik; Kim, Eung Soo; Yeom, Keum Soo [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

2002-03-15

The objective of the present research is to perform the separate effect tests and to assess the RELAP5/MOD3.2 code for the analysis of thermal-hydraulic behavior in the reactor coolant system and the improvement of the auditing technology of safety analysis. Three Separate Effect Tests (SETs) are the reflux condensation in the U-tube, the direct contact condensation in the hot-leg and the mixture level buildup in the pressurizer. The experimental data and the empirical correlations are obtained through SETs. On the ases of the three SET works, models in RELAP5 are modified and improved, which are compared with the data. The Korea Standard Nuclear Power Plant (KSNP) are assessed using the modified RELAP5. In the reflux condensation test, the data of heat transfer coefficients and flooding are obtained and the condensation models are modified using the non-iterative model, as results, modified code better predicts the data. In the direct contact condensation test, the data of heat transfer coefficients are obtained for the cocurrent and countercurrent flow between the mixture gas and the water in condition of horizontal stratified flow. Several condensation and friction models are modified, which well predict the present data. In the mixture level test, the data for the mixture level and the onset of water draining into the surge line are obtained. The standard RELAP5 over-predicts the mixture level and the void fraction in the pressurizer. Simple modification of model related to the pool void fraction is suggested. The KSNP is assessed using the standard and the modified RELAP5 resulting from the experimental and code works for the SETs. In case of the pressurizer manway opening with available secondary side of the steam generators, the modified code predicts that the collapsed level in the pressurizer is little accumulated. The presence and location of the opening and the secondary condition of the steam generators have an effect on the coolant inventory. The

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

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

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

A thermal analysis study on proprietary quick-setting mixed binder system

International Nuclear Information System (INIS)

Deneanu, N.; Dulama, M.; Diaconescu, C.

2013-01-01

The present paper presents a study, by thermal analysis, of the hydration of a mixed binder system consist of Portland composite cement type CEM II/A-M(S-LL) 32,5 R and calcium aluminate cement type GORKAL 70, patented by the authors for embedding spent radioactive solvents. Cement hydration reactions are often affected by the presence of a spent radioactive solvent, usually retarded. This process can be monitored by thermal analysis and often quantified. The data are being used as a reference to compare the changes in the cement blend hydration during the solidification caused by the presence of the spent radioactive solvents itself or by emulsifier additives which are being availed. Detailed analysis of DTG (derivative thermo gravimetric) decomposition profiles of portlandite and carbonate enabled the evaluation of admixture-related parameters concerning portlandite formation and also indicated the behaviour of specific carbonates during the hydration process. (authors)

SSTL Based Low Power Thermal Efficient WLAN Specific 32bit ALU Design on 28nm FPGA

DEFF Research Database (Denmark)

Kalia, Kartik; Pandey, Bishwajeet; Das, Teerath

2016-01-01

at minimum and maximum temperature as compared to all other considered I/O standards. This design has application where 32bit ALU design is considered for designing an electronic device such as WLAN. The design can be implemented on different nano chips for better efficiency depending upon the design...... with consideration of airflow toward hit sink and different frequency on which ALU operate in network processor or any WLAN devices. We have done total power analysis of WLAN operating on different frequencies. We have considered a set of frequencies, which are based on IEEE 802.11 standards. First we did...... efficient IO standard. While analyzing we found out that when WLAN device shift from 343.15K to 283.15K, there is maximum thermal power reduction in SSTL135_R as compared to all considered I/O standards. When we compared same I/Os for different frequencies we observed maximum thermal efficiency in SSTL15...

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....

Ni(NH32(NO32—A 3-D Network through Bridging Nitrate Units Isolated from the Thermal Decomposition of Nickel Hexammine Dinitrate

Directory of Open Access Journals (Sweden)

Joachim Breternitz

2018-06-01

Full Text Available Nickel nitrate diammine, Ni(NH32(NO32, can be synthesised from the thermal decomposition of nickel nitrate hexammine, Ni[(NH36](NO32. The hexammine decomposes in two distinct major stages; the first releases 4 equivalents of ammonia while the second involves the release of NOx, N2, and H2O to yield NiO. The intermediate diammine compound can be isolated following the first deammoniation step or synthesised as a single phase from the hexammine under vacuum. Powder X-ray diffraction (PXD experiments have allowed the structure of Ni(NH32(NO32 to be solved for the first time. The compound crystallises in orthorhombic space group Pca21 (a = 11.0628 (5 Å, b = 6.0454 (3 Å, c = 9.3526 (4 Å; Z = 4 and contains 11 non-hydrogen atoms in the asymmetric unit. Fourier transform infrared (FTIR spectroscopy demonstrates that the bonding in the ammine is consistent with the structure determined by PXD.

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.

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

Evaluation of the effects of break nozzle configuration in the Semiscale Mod-1 system

International Nuclear Information System (INIS)

Hanson, R.G.

1977-08-01

The Semiscale Mod-1 Program has utilized two different break nozzle configurations in the test system. An evaluation has been made to determine the effect these break nozzle configurations have on system thermal-hydraulic response during a 200 percent double-ended cold leg break loss-of-coolant accident simulation. The first nozzle was a convergent-divergent nozzle (Henry nozzle) and the second, an elongated constant area throat nozzle. Analysis is confined primarily to system response phenomena observed to be affected by the nozzle configuration and concentrates on the fluid response at the break and the resulting core behavior during subcooled and saturated blowdown. The evaluation shows that considerable difference in system response occurs as a result of the difference in break nozzle configuration. The elongated throat nozzle was scaled from the Loss-of-Fluid Test (LOFT) nozzle geometry and since the LOFT counterpart tests were designed to provide results for the LOFT Program, the elongated throat nozzle was used in the subsequent LOFT counterpart tests

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)

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

Mod-5A Wind Turbine Generator Program Design Report. Volume 4: Drawings and Specifications, 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. Volume 4 contains the drawings and specifications that were developed in preparation for building the MOD-5A wind turbine generator. This is the first of five books of volume four. It contains structural design criteria, generator step-up transformer specs, specs for design, fabrication and testing of the system, specs for the ground control enclosure, systems specs, slip ring specs, and control system specs.

Thermal energy systems design and analysis

CERN Document Server

Penoncello, Steven G

2015-01-01

IntroductionThermal Energy Systems Design and AnalysisSoftwareThermal Energy System TopicsUnits and Unit SystemsThermophysical PropertiesEngineering DesignEngineering EconomicsIntroductionCommon Engineering Economics NomenclatureEconomic Analysis Tool: The Cash Flow DiagramTime Value of MoneyTime Value of Money ExamplesUsing Software to Calculate Interest FactorsEconomic Decision MakingDepreciation and TaxesProblemsAnalysis of Thermal Energy SystemsIntroductionNomenclatureThermophysical Properties of SubstancesSuggested Thermal Energy Systems Analysis ProcedureConserved and Balanced QuantitiesConservation of MassConservation of Energy (The First Law of Thermodynamics)Entropy Balance (The Second Law of Thermodynamics)Exergy Balance: The Combined LawEnergy and Exergy Analysis of Thermal Energy CyclesDetailed Analysis of Thermal Energy CyclesProblemsFluid Transport in Thermal Energy SystemsIntroductionPiping and Tubing StandardsFluid Flow FundamentalsValves and FittingsDesign and Analysis of Pipe NetworksEconomi...

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....

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

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.

Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty; Qualificacao e aplicacao de codigo de acidentes de reatores nucleares com capacidade interna de avaliacao de incerteza

Energy Technology Data Exchange (ETDEWEB)

Borges, Ronaldo Celem

2001-10-15

This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

Fast Wave Transmission Measurements on Alcator C-Mod

Science.gov (United States)

Reardon, J.; Bonoli, P. T.; Porkolab, M.; Takase, Y.; Wukitch, S. J.

1997-11-01

Data are presented from an array of single-turn loop probes newly installed on the inner wall of C-Mod, directly opposite one of the two fast-wave antennas. The 8-loop array extends 32^circ in the toroidal direction at the midplane and can distinguish electromagnetic from electrostatic modes. Data are acquired by 1GHz digitizer, spectrum analyzer, and RF detector circuit. Phase measurements during different heating scenarios show evidence of both standing and travelling waves. The measurement of toroidal mode number N_tor (conserved under the assumption of axisymmetry) is used to guide the toroidal full-wave code TORIC(Brambilla, M., IPP Report 5/66, February 1996). Amplitude measurements show modulation both by Type III ELMs and sawteeth; the observed sawtooth modulation may be interpreted as due to changes in central absorption. The amplitude of tildeB_tor measured at the inner wall is compared to the prediction of TORIC.

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

Using computer program RELAP5/MOD2 on microcomputers

International Nuclear Information System (INIS)

Grgic, D.; Bajs, T; Cavlina, N.; Debrecin, N.

1990-01-01

Our work on installation of RELAP5/MOD2 code on IBM4341, mVAX 11, MGT-386 and COMPAQ-386/20e computers is described. Main characteristics of RELAP5/MOD2 structure programming style and differences between FORTRAN VS, VAX-11 FORTRAN and NDP FORTRAN 386 are presented. We discussed basic philosophy used in modification and testing and test results. (author)

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.

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)

Evaluation of the RELAP5/MOD3 multidimensional component model

International Nuclear Information System (INIS)

Tomlinson, E.T.; Rens, T.E.; Coffield, R.D.

1994-01-01

Accurate plenum predictions, which are directly related to the mixing models used, are an important plant modeling consideration because of the consequential impact on basic transient performance calculations for the integrated system. The effect of plenum is a time shift between inlet and outlet temperature changes to the particular volume. Perfect mixing, where the total volume interacts instantaneously with the total inlet flow, does not occur because of effects such as inlet/outlet nozzle jetting, flow stratification, nested vortices within the volume and the general three-dimensional velocity distribution of the flow field. The time lag which exists between the inlet and outlet flows impacts the predicted rate of temperature change experienced by various plant system components and this impacts local component analyses which are affected by the rate of temperature change. This study includes a comparison of two-dimensional plenum mixing predictions using CFD-FLOW3D, RELAP5/MOD3 and perfect mixing models. Three different geometries (flat, square and tall) are assessed for scalar transport times using a wide range of inlet velocity and isothermal conditions. In addition, the three geometries were evaluated for low flow conditions with the inlet flow experiencing a large step temperature decrease. A major conclusion from this study is that the RELAP5/MOD3 multidimensional component model appears to be adequately predicting plenum mixing for a wide range of thermal-hydraulic conditions representative of plant transients

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

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

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

neutron radiography. Report prepared from contributions by members of the MOD Working Party on Neutron Radiography

International Nuclear Information System (INIS)

Halmshaw, R.

1977-03-01

Radiography with thermal or cold neutrons has some special advantages over X-rays and gamma rays, and some facilities for neutron radiography exist in the Ministry of Defence. This report gives a brief and simple description of the technique, its advantages and disadvantages, and is illustrated with a number of Ordnance applications taken from MOD work, to show examples where neutron radiographs provided extra important information not available from X- or gamma radiography. The facilities available in the UK for neutron radiography are listed. (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.

RELAP5 / MOD3.2 analysis of INSC standard problem INSCSP - V4 : investigation of heat transfer for partly uncovered VVER-1000 core at the test facility KS (RRC K1)

International Nuclear Information System (INIS)

Tentner, A.; Ahrens, J. W.

2002-01-01

The RELAP5/MOD3.2 computer program has been used to analyze a series of tests investigating heat-transfer from a partly uncovered VVER-1000 core in the KS test facility at the Russian Research Center ''Kurchatov Institute'' (RRC-KI). The analysis documented represents VVER Standard Problem 4 in Joint Project 6, which is the investigation of Computer Code Validation for Transient Analysis of RBMK and VVER Reactors, between the United States and Russian International Nuclear Safety Centers. The experiment facility and data, RELAP5 nodalization, and results are shown for one of the six tests defined in Standard Problem 4. Only part of the data was analyzed due to our conclusion that the available experimental data is not sufficient to allow the modeling of the actual experiment sequence. The experiment initial conditions were reached through a series of transient processes, about which no quantitative information was available. This has required the modeling of an arbitrary computational transient, with the goal of reaching initial conditions similar to those observed during the experiment. The use of an arbitrary transient introduces many degrees of freedom in the analysis, i.e. initial computational values that influence the entire sequence of events, including the loop behavior during the experiment time window. Reasonable agreement between RELAP5 and the experiment data can be obtained by manipulating a number of initial computational values, including the liquid level in the fuel assembly model, the liquid level in the annular region, the quality of the saturated vapor in the voided loop regions, etc. Our study has focused on exploring the sensitivity of results to changes in these initial values which are not based on experimental information, but are selected with the goal of matching the experimentally observed behavior during the experiment time window. We have shown that changes in several initial arbitrary values can lead to similar changes in the

RELAP5/MOD2 assessment at Babcock and Wilcox

International Nuclear Information System (INIS)

Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Turk, C.

1986-01-01

Babcock and Wilcox (B and W) has been working with the code developers at EG and G Idaho, Inc. and the NRC assessing the RELAP5/MOD2 computer code by simulating selected separate effects tests. The purpose of this B and W Owners Group-sponsored assessment was to evaluate RELAP5/MOD2 for use in design calculations for the MIST and OTIS integral system tests and in predicting pressurized water reactor (PWR) transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (Cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve specific predictive capabilities of RELAP5/MOD2

Assessment of the RELAP4/MOD6 thermal-hydraulic transient code for PWR experimental applications. Addendum. Analyses completed and reported in FY 1979. Interim report

International Nuclear Information System (INIS)

1979-09-01

The results of three subtasks that complete the assessment of the RELAP4/MOD6 computer code are reported. These subtasks constitute the remainder of a broadly scoped assessment matrix defined and described in detail in a previously published document. The specific subtasks provide comparisons of code calculations with experimental results from core blowdown and critical-flow separate-effects experiments and from an integral systems-effects loss-of-coolant experiment. The basic emphasis of the comparisons is in the presentation of the study results in error form suitable for statistical analysis

Ohmic ITBs in Alcator C-Mod

Science.gov (United States)

Fiore, C. L.; Rowan, W. L.; Dominguez, A.; Hubbard, A. E.; Ince-Cushman, A.; Greenwald, M. J.; Lin, L.; Marmar, E. S.; Reinke, M.; Rice, J. E.; Zhurovich, K.

2007-11-01

Internal transport barrier plasmas can arise spontaneously in ohmic Alcator C-Mod plasmas where an EDA H-mode has been developed by magnetic field ramping. These ohmic ITBs share the hallmarks of ITBs created with off-axis ICRF injection in that they have highly peaked density and pressure profiles and the peaking can be suppressed by on-axis ICRF. There is a reduction of particle and thermal flux in the barrier region which then allows the neoclassical pinch to peak the central density. Recent work on ITB onset conditions [1] which was motivated by turbulence studies [2] points to the broadening of the Ti profile with off-axis ICRF acting to reduce the ion temperature gradient. This suppresses ITG instability driven particle fluxes, which is thought to be the primary mechanism for ITB formation. The object of this study is to examine the characteristics of ohmic ITBs to find whether the stability of plasmas and the plasma parameters support the onset model. [1]K. Zhurovich, et al., To be published in Nuclear Fusion [2] D. R. Ernst, et al., Phys. Plasmas 11, 2637 (2004)

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.

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...

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...

Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

International Nuclear Information System (INIS)

D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

2003-01-01

A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

Evaluation of a loss of residual heat removal event during mid-loop operation

International Nuclear Information System (INIS)

Seul, Kwang Won; Bang, Young Seok; Lee, Sukho; Kim, Hho Jung

1996-01-01

The potential for the RELAP5/MOD3.2 was assessed for the loss-of -RHR event during the mid-loop operation and the predictability of major thermal-hydraulic phenomena was also evaluated for the long term transient. The analysis results of the typical two cases(cold leg opening case and pressurizer opening case) were compared with experimental data which was conducted at ROSA-IV/LSTF in Japan. As a result, it was shown that the code was capable of simulating the thermal-hydraulic transport process with appropriate time step during the reduced inventory operation with the loss-of-RHR system

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

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,

Parameterization and Observability Analysis of Scalable Battery Clusters for Onboard Thermal Management Paramétrage et analyse d’observabilité de clusters de batteries de taille variable pour une gestion thermique embarquée

Directory of Open Access Journals (Sweden)

Lin Xinfan

2013-03-01

Full Text Available Although the battery surface temperature is commonly measured, the core temperature of a cell may be much higher hence more critical than the surface temperature. The core temperature of a battery, though usually unmeasured in commercial applications, can be estimated by an observer, based on a lumped-parameter battery thermal model and the measurement of the current and the surface temperature. Even with a closed loop observer based on the measured surface temperature, the accuracy of the core temperature estimation depends on the model parameters. For such purpose, an online parameterization methodology and an adaptive observer are designed for a cylindrical battery. The single cell thermal model is then scaled up to create a battery cluster model to investigate the temperature pattern of the cluster. The modeled thermal interconnections between cells include cell to cell heat conduction and convection to the surrounding coolant flow. An observability analysis is performed on the cluster before designing a closed loop observer for the pack. Based on the analysis, guidelines for determining the minimum number of required sensors and their exact locations are derived that guarantee the observability of all temperature states. Bien que la température de surface d’une batterie soit généralement mesurée, la température interne d’une cellule peut être beaucoup plus élevée donc plus critique que la température de surface. La température interne d’une batterie, pourtant normalement non mesurée dans les applications commerciales, peut être évaluée par un observateur, sur la base d’un modèle thermique de batterie à constantes localisées et à partir de la mesure du courant et de la température de surface. Même avec un observateur en boucle fermée basé sur la température de surface mesurée, la précision de l’estimation de la température interne dépend des constantes du modèle. Dans cette optique, une méthodologie de

Incorporation of improvements in the intercomparison ISP.38 exercise

International Nuclear Information System (INIS)

Carlos, S.; Martorell, S.; Serradell, V.; Verdu, G.; Rodenas, J.; Saura, J.; Villanueva, J. F.; Miro, R.

1998-01-01

Nowadays, several on transient simulations under low power and shutdown conditions are being performed at the Chemical and Nuclear Engineering Department of the Polytechnic University of Valencia, in collaboration with the Spanish Nuclear Safety Council. One of the activities involved in this study was the participation in the International Standard Problem (ISP) 38, which consisted of simulating the loss the residual heat removal system in the BETHSY installation. The main objective of the ISP-38 was the comparison of the thermal-hydraulic codes capability to simulate shutdown conditions. Once the ISP was finished, several sensitivity studies were performed. These studies referred to the analysis of the results obtained using different plant nodalization, and using different versions of the RELAP5. The calculations were submitted with RELAP5/MOD3.2 and with the RELAP5/MOD3.2 fg development version. Although after the sensitivity study the results obtained were closer to the experimental ones than before, more improvements may be introduced with other sensitivity studies. (Author)

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

Science.gov (United States)

1984-01-01

The design, development and analysis of the 7.3 MW MOD-5A wind turbine generator are 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 volume contains 5 books of which this is the fourth, providing drawings 47A380128 through 47A387125. In addition to the parts listing and where-used list, the logic design of the controller software and the code listing of the controller software are provided. Also given are the aerodynamic profile coordinates.

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.

Comparisons of TRAC-PF-1 calculations with semiscale Mod-3 small-break tests S-SB-P1 and S-SB-P7

International Nuclear Information System (INIS)

Sahota, M.S.

1982-01-01

Semiscale Tests S-SB-P1 and S-SB-P7 conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1). The results are used to assess TRAC-PF1 predictions of thermal-hydraulic phenomena and the effects of break size and pump operation on system response during slow transients. Tests S-SB-P1 and S-SB-P7 simulated an equivalent pressurized-water-reactor (PWR) 2.5% communicative cold-leg break for early and late pump trips, respectively, with only high-pressure injection (HPI) into the cold legs. The parameters examined include break flow, primary-system pressure response, primary-system mass distribution, and core characteristics

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)

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

Thermal analysis of annular fins with temperature-dependent thermal properties

Institute of Scientific and Technical Information of China (English)

I. G. AKSOY

2013-01-01

The thermal analysis of the annular rectangular profile fins with variable thermal properties is investigated by using the homotopy analysis method (HAM). The thermal conductivity and heat transfer coefficient are assumed to vary with a linear and power-law function of temperature, respectively. The effects of the thermal-geometric fin parameter and the thermal conductivity parameter variations on the temperature distribution and fin efficiency are investigated for different heat transfer modes. Results from the HAM are compared with numerical results of the finite difference method (FDM). It can be seen that the variation of dimensionless parameters has a significant effect on the temperature distribution and fin efficiency.

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)

Application of RELAP/SCDAPSIM to TRIGA system thermal hydraulic analysis

International Nuclear Information System (INIS)

Allison, C.M.; Hohorst, J.K.; Quamrul Huda, Md.

2010-01-01

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 the international SCDAP 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 publicly available RELAP5 and SCDAP models in combination with advanced programming and numerical techniques and other SDTP-member modeling/user options. This paper describes the development of a representative input model for the 3MW TRIGA research reactor at AERE Bangladesh, describes the testing and qualification of the model using MOD4.0 advanced input checking and graphical display options, and then presents representative results for selected calculations. (author)

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.

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.

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...

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

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'...

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

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)

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

Condensation heat transfer coefficient with noncondensible gases for heat transfer in thermal hydraulic codes

International Nuclear Information System (INIS)

Banerjee, S.; Hassan, Y.A.

1995-01-01

Condensation in the presence of noncondensible gases plays an important role in the nuclear industry. The RELAP5/MOD3 thermal hydraulic code was used to study the ability of the code to predict this phenomenon. Two separate effects experiments were simulated using this code. These were the Massachusetts Institute of Technology's (MIT) Pressurizer Experiment, the MIT Single Tube Experiment. A new iterative approach to calculate the interface temperature and the degraded heat transfer coefficient was developed and implemented in the RELAP5/MOD3 thermal hydraulic code. This model employs the heat transfer simultaneously. This model was found to perform much better than the reduction factor approach. The calculations using the new model were found to be in much better agreement with the experimental values

Condensation heat transfer coefficient with noncondensible gases for heat transfer in thermal hydraulic codes

Energy Technology Data Exchange (ETDEWEB)

Banerjee, S.; Hassan, Y.A. [Texas A& M Univ., College Station, TX (United States)

1995-09-01

Condensation in the presence of noncondensible gases plays an important role in the nuclear industry. The RELAP5/MOD3 thermal hydraulic code was used to study the ability of the code to predict this phenomenon. Two separate effects experiments were simulated using this code. These were the Massachusetts Institute of Technology`s (MIT) Pressurizer Experiment, the MIT Single Tube Experiment. A new iterative approach to calculate the interface temperature and the degraded heat transfer coefficient was developed and implemented in the RELAP5/MOD3 thermal hydraulic code. This model employs the heat transfer simultaneously. This model was found to perform much better than the reduction factor approach. The calculations using the new model were found to be in much better agreement with the experimental values.

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

International Nuclear Information System (INIS)

Wu, Pan; Xiong, Xiaofei; Shan, Jianqiang; Gou, Junli; Zhang, Bin; Zhang, Bo

2016-01-01

Highlights: • A new heat transfer package has been developed. • It has been incorporated into RELAP5/MOD3.3 to verify its advantages. • The results of modified code were compared with available experimental data. • The results showed that higher prediction accuracy was achieved. - Abstract: The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

Energy Technology Data Exchange (ETDEWEB)

Wu, Pan; Xiong, Xiaofei; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli; Zhang, Bin; Zhang, Bo

2016-12-15

Highlights: • A new heat transfer package has been developed. • It has been incorporated into RELAP5/MOD3.3 to verify its advantages. • The results of modified code were compared with available experimental data. • The results showed that higher prediction accuracy was achieved. - Abstract: The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

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

Development of a multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3 and its verification

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-06-01

A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other

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...

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)

Current and anticipated uses of thermal hydraulic codes in Korea

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung-Doo; Chang, Won-Pyo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-07-01

In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.

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

Application of RELAP5/MOD3.1 code to the LOFT test L3-6

International Nuclear Information System (INIS)

Pylev, S.S.; Roginskaja, V.L.

1998-02-01

A calculation of LOFT Experiment L3-6, a small break equivalent to a 4-in diameter rupture in the cold leg of a four-loop commercial pressurized water reactor, has been performed to help validate RELAP5/MOD3.1 for this application. The version of the code to be used is SCDAP/RELAP5/MOD3.1.8d0. Three calculations were carried out in order to study the sensitivity to change break nozzle superheated discharge coefficient. Conducted comparative analysis of the LOFT L3-6 experiment shows on the whole a reasonable agreement between calculated data. Some discrepancies in the system pressure do not distort a picture of the transient. 6 refs

Application of RELAP5/MOD3.1 code to the LOFT test L3-6

Energy Technology Data Exchange (ETDEWEB)

Pylev, S.S.; Roginskaja, V.L.

1998-02-01

A calculation of LOFT Experiment L3-6, a small break equivalent to a 4-in diameter rupture in the cold leg of a four-loop commercial pressurized water reactor, has been performed to help validate RELAP5/MOD3.1 for this application. The version of the code to be used is SCDAP/RELAP5/MOD3.1.8d0. Three calculations were carried out in order to study the sensitivity to change break nozzle superheated discharge coefficient. Conducted comparative analysis of the LOFT L3-6 experiment shows on the whole a reasonable agreement between calculated data. Some discrepancies in the system pressure do not distort a picture of the transient. 6 refs.

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

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.

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 ...

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

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. .

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-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.

Thermal Hydraulic Assessment for Loss of SDCS Event During the Outage of CANDU Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Jonghyun [Gnest, Inc. Taejon (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

2006-07-01

During the outage(overhaul) of the nuclear power plant, there are several operating states other than the full power state, that is 'Hot-Zero Power', 'Depressurized-Cooldown', and 'Partially Drained'. Until now safety assessment has not been done much for this operating state of CANDU type reactor worldwide. For the accuracy and confidence of PSA for the CANDU outage, the safety analysis is necessary. At the first stage, we analyzed the thermal hydraulic characteristics and safety of the postulated event of loss of shutdown cooling system (SDCS) during the partially drained state which is the longest one in the middle of outage period. As an analysis tool, this study uses the best estimate thermal hydraulic code, RELAP5/CANDU which was modified according to the CANDU specific characteristics and based on RELAP5.Mod3.

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)

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

Modelling disassembled fuel bundles using CATHENA MOD-3.5a under LOCA/LOECC conditions

Energy Technology Data Exchange (ETDEWEB)

Lei, Q M; Sanderson, D B; Dutton, R [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1996-12-31

CATHENA MOD-3.5a is a multipurpose thermalhydraulic computer code developed primarily to analyse postulated loss-of-coolant scenarios for CANDU nuclear reactors. The code contains a generalized heat transfer package that enables it to model the behaviour of a fuel channel in great detail. Throughout the development of the CATHENA code, considerable effort has been devoted to evaluating, validating and documenting its overall capability as a design and safety assessment tool. Specific attention has focused on its ability to predict fuel channel behaviour under postulated accident conditions. This paper describes an investigation of CATHENA`s ability to predict the thermal-chemical responses of a fuel channel in which the 37-element bundles were assumed to disassemble and rearrange into a closed-packed stack of elements at the bottom of the pressure tube. A representative disassembled bundle geometry was modelled during a simulated loss-of-coolant accident scenario using CATHENA MOD-3.5a/Rev 0, with superheated steam being the only coolant available. Thermal conduction in the radial and circumferential directions was calculated for individual fuel elements, the pressure tube, and the calandria tube. Radiation view factors for the intact and disassembled bundle geometries were calculated using a CATHENA utility program. Inter-element metal-to-metal contact was accounted for using the CATHENA solid-solid contact model. An offset pressure-tube configuration, representing a partially sagged pressure tube, and the effect of steam starvation on the exothermic zirconium-steam reaction, were included in the CATHENA model. The CATHENA-predicted results show a dramatic suppression of heat generation from the zirconium-steam reaction when bundle disassembly is initiated. The predicted results show a smaller temperature increase in the fuel sheaths and the pressure tube for the disassembled bundle geometry, compared to the temperature excursion for the intact bundle. (author

Insertion of lead lithium eutectic mixture in RELAP/SCDAPSIM Mod 4.0 for Fusion Reactor Systems

International Nuclear Information System (INIS)

Tiwari, Ashutosh; Allison, Brian; Hohorst, J.K.; Wagner, R.J.; Allison, Chris

2012-01-01

Highlights: ► Thermodynamic and transport properties of lead lithium eutectic mixture have been inserted in RELAP/SCDAPSIM MOD 4.0 code. ► Code results are verified for a simple pipe problem with lead lithium eutectic mixture flowing in it. ► Code is calculating the inserted properties of lead lithium eutectic mixture to a fairly good agreement. - Abstract: RELAP/SCDAPSIM Mod 4.0 code was developed by Innovative System Software (ISS) for the analysis of nuclear power plants (NPPs) cooled by light water and heavy water. Later on the code was expanded to analyze the NPPs cooled by liquid metal, in this sequence: lead bismuth eutectic mixture, liquid sodium and lead lithium eutectic mixture (LLE) are inserted in the code. This paper focuses on the insertion of liquid LLE as a coolant for NPPs in the RELAP/SCDAPSIM Mod 4.0 code. Evaluation of the code was made for a simple pipe problem connected with heat structures having liquid LLE as a coolant in it. The code is predicting well all the thermodynamic and transport properties of LLE.

Application of Thermal Network Model to Transient Thermal Analysis of Power Electronic Package Substrate

Directory of Open Access Journals (Sweden)

Masaru Ishizuka

2011-01-01

Full Text Available In recent years, there is a growing demand to have smaller and lighter electronic circuits which have greater complexity, multifunctionality, and reliability. High-density multichip packaging technology has been used in order to meet these requirements. The higher the density scale is, the larger the power dissipation per unit area becomes. Therefore, in the designing process, it has become very important to carry out the thermal analysis. However, the heat transport model in multichip modules is very complex, and its treatment is tedious and time consuming. This paper describes an application of the thermal network method to the transient thermal analysis of multichip modules and proposes a simple model for the thermal analysis of multichip modules as a preliminary thermal design tool. On the basis of the result of transient thermal analysis, the validity of the thermal network method and the simple thermal analysis model is confirmed.

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...

Conversion of the thermal hydraulics components of Almaraz NPP model from RELAP5 into TRAC-M

International Nuclear Information System (INIS)

Queral, C.; Mulas, J.; Collazo, I.; Concejal, A.; Burbano, N.; Gallego, I.; Lopez Lechas, A.

2002-01-01

In the scope of a joint project between the Spanish Nuclear Regulatory Commission (CSN) and the electric energy industry of Spain (UNESA) on the USNRC state-of-the-art thermal hydraulic code, TRAC-M, there is a task relating to the translation of the Spanish NPP models from other TH codes to the new one. As part of this project, our team is working on the translation of Almaraz NPP model from RELAP5/MOD3.2 to TRAC-M. At present, several portions of the input deck have been converted to TRAC-M, and the output data have also been compared with RELAP5 data. This paper refers to the translation of the following thermal hydraulic models: pressurizer, hot and cold legs (including the pumps and the injection systems), and steam generators. The comparison of the results obtained with both codes shows a good agreement. However, some difficulties have been found in the translation of some code components, like pipes, heat structures, pumps, branchs, valves and boundary conditions. In this paper, these translation problems and their solutions are described.(author)

VIPRE-01: A thermal-hydraulic code for reactor cores

International Nuclear Information System (INIS)

Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.; Nomura, K.K.

1989-08-01

The VIPRE-01 thermal hydraulics code for PWR and BWR analysis has undergone significant modifications and error correction. This manual for the updated code, designated as VIPRE-01 Mod-02, describes improvements that eliminate problems of slow convergence with the drift flux model in transient simulation. To update the VIPRE-01 code and its documentation the drift flux model of two-phase flow was implemented and error corrections developed during VIPRE-01 application were included. The project team modified the existing VIPRE-01 equations into drift flux model equations by developing additional terms. They also developed and implemented corrections for the errors identified during the last four years. They then validated the modified code against standard test data using selected test cases. The project team prepared documentation revisions reflecting code improvements and corrections to replace the corresponding sections in the original VIPRE documents. The revised VIPRE code, designated VIPRE-01 Mod-02, incorporates improvements that eliminate many shortcomings of the previous version. During the validation, the code produced satisfactory output compared with test data. The revised documentation is in the form of binder pages to replace existing pages in three of the original manuals

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

Energy Technology Data Exchange (ETDEWEB)

Li, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Qiu, Suizheng; Su, Guanghui; Jiao, Hong [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Bai, Yunqing; Chen, Hongli [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Yican, E-mail: yican.Wu@Fds.Org.Cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

2013-06-15

Highlights: • Thermal-hydraulic and safety analysis on DFLL-TBM system is performed. • The TBM FW maximum temperature is 541 °C under steady state condition. • The TBM FW maximum temperature does not exceed the melt point of CLAM steel 1500 °C. • Neither the VV pressurization nor vault pressure build-up goes beyond 0.2 MPa. -- Abstract: China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current

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...

Du teikei à l’AMAP, un modèle acculturé

Directory of Open Access Journals (Sweden)

Jean Lagane

2011-05-01

Full Text Available Cet article compare à travers deux ethnographies en France et au Japon le modèle français de l’AMAP – association de maintien de l’agriculture paysanne – au modèle dont il se réclame, le teikei, système japonais de partenariat agricole entre producteurs et consommateurs. Terrain en mutation permanente, le développement des formes de circuit court et leurs emprunts accentuent lors de la transposition la nécessité de coller le plus étroitement possible à la réalité du terrain, à son histoire sociétale, à ses traditions institutionnelles et culturelles, à ses crises, à ses dynamiques. L’analyse fait apparaître à travers les similitudes, dissemblances et adaptations de ce modèle d’emprunt, l’incidence et la complexité de la prise en compte de facteurs culturels.After two ethnological fieldworks completed in France and Japan, this article deals with the French system of Community Supported Agriculture - AMAP - and its Japanese inspiration counterpart, the Teikei system built upon local solidarity-based partnerships between farmers and members. Constantly evolving, the growing process of agriculture one-shot sale circuits and the way they are received in other cultures recalls the necessity to tune with the reality of these societies, their history, their institutional and cultural traditions as well as their crises and dynamics. The analysis emphasizes the complexity of the acculturation process related with points of convergence and discrepancy.

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...

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)

Investigation of the critical edge ion heat flux for L-H transitions in Alcator C-Mod and its dependence on B T

Science.gov (United States)

Schmidtmayr, M.; Hughes, J. W.; Ryter, F.; Wolfrum, E.; Cao, N.; Creely, A. J.; Howard, N.; Hubbard, A. E.; Lin, Y.; Reinke, M. L.; Rice, J. E.; Tolman, E. A.; Wukitch, S.; Ma, Y.; ASDEX Upgrade Team; Alcator C-Mod Team

2018-05-01

This paper presents investigations on the role of the edge ion heat flux for transitions from L-mode to H-mode in Alcator C-Mod. Previous results from the ASDEX Upgrade tokamak indicated that a critical value of edge ion heat flux per particle is needed for the transition. Analysis of C-Mod data confirms this result. The edge ion heat flux is indeed found to increase linearly with density at given magnetic field and plasma current. Furthermore, the Alcator C-Mod data indicate that the edge ion heat flux at the L-H transition also increases with magnetic field. Combining the data from Alcator C-Mod and ASDEX Upgrade yields a general expression for the edge ion heat flux at the L-H transition. These results are discussed from the point of view of the possible physics mechanism of the L-H transition. They are also compared to the L-H power threshold scaling and an extrapolation for ITER is given.

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....

Determination of phosphorous in cannabis by neutron activation analysis - measurement of 32P Cerenkov radiation by liquid scintillaton spectrometer

International Nuclear Information System (INIS)

Shinogi, M.; Mori, I.

1977-05-01

Thermal neutron activaton analysis with measurement of 32 P Cerenkov radiation by liquid scintillation spectrometer was used to determine phosphorus in cannabis. After irradiation of the sample, wet ashing was carried out with conc. nitric acid and 70% perchloric acid. The solution in l M perchloric acid transferred to an inorganic ion-exchange column containing acid aluminium oxide and phosphorus was quantitatively eluted with 1M hydrofluoric acid. The 32 P radioactivity of each fraction of the eluate was counted with Cerenkov radiation by a liquid scintillation spectrometer from 2 to 7 weeks after the irradiation. The activity curve decayed with 32 P half-life. The isotope channel ratio technique was applied for the quench correction. The optimal experimental conditions for chemical separation of phosphorus and for measuring the 32 P Cerenkov radiation were also examined. (Author)

Experiment data report for Semiscale Mod-1 test S-02-5 (blowdown heat transfer test)

International Nuclear Information System (INIS)

1975-12-01

Recorded test data are presented for Test S-02-5 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-5 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 water-cooled nuclear reactor system and to provide data for the assessment of the Loss-of-Fluid Test (LOFT) design basis. Test S-02-5 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,253 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with full core power (1.6 MW). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was set to achieve the full design core temperature differential of 66 0 F. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling occurs

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

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.

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

Thermal Analysis of Disposal of High-Level Nuclear Waste in a Generic Bedded Salt repository using the Semi-Analytical Method.

Energy Technology Data Exchange (ETDEWEB)

Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-05-01

An example case is presented for testing analytical thermal models. The example case represents thermal analysis of a generic repository in bedded salt at 500 m depth. The analysis is part of the study reported in Matteo et al. (2016). Ambient average ground surface temperature of 15°C, and a natural geothermal gradient of 25°C/km, were assumed to calculate temperature at the near field. For generic salt repository concept crushed salt backfill is assumed. For the semi-analytical analysis crushed salt thermal conductivity of 0.57 W/m-K was used. With time the crushed salt is expected to consolidate into intact salt. In this study a backfill thermal conductivity of 3.2 W/m-K (same as intact) is used for sensitivity analysis. Decay heat data for SRS glass is given in Table 1. The rest of the parameter values are shown below. Results of peak temperatures at the waste package surface are given in Table 2.

20% inlet header break analysis of Advanced Heavy Water Reactor

International Nuclear Information System (INIS)

Srivastava, A.; Gupta, S.K.; Venkat Raj, V.; Singh, R.; Iyer, K.

2001-01-01

The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

Posttest analysis of MIST Test 330302 using TRAC-PF1/MOD1

International Nuclear Information System (INIS)

Boyack, B.E.

1992-09-01

This report discusses a posttest analysis of Multi-Loop Integral System Test (MIST) 330302 which has been performed using TRAC-PF1/MOD1. This test was one of group performed in the MIST facility to investigate high-pressure injection (HPI)-power-operated relief valve (PORV) cooling, also known as feed-and-bleed cooling. In Test 330302, HPI cooling was delayed 20 min after opening and locking the PORV open to induce extensive system voiding. We have concluded that the TRAC-calculated results are in reasonable overall agreement with the data for Test 330302. All major trends and phenomena were correctly predicted. Differences observed between the measured and calculated results have been traced and related, in part, to deficiencies in our knowledge of the facility configuration and operation. We have identified two models forwhich additional review is appropriate. However, in general, the TRAC closure models and correlations appear to be adequate for the prediction of the phenomena expected to occur during feed-and-bleed transientsin the MIST facility. We believe that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications. Conclusions reached regarding use of the code to calculate similar phenomena in full-size plants (scaling implications) and regulatory implications of this work are also presented

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.

Topology and transport in the edge region of RFX-mod helical regimes

International Nuclear Information System (INIS)

Scarin, P.; Vianello, N.; Agostini, M.; Spizzo, G.; Spolaore, M.; Zuin, M.; Cappello, S.; Carraro, L.; Cavazzana, R.; De Masi, G.; Martines, E.; Moresco, M.; Munaretto, S.; Puiatti, M. E.; Valisa, M.

2011-01-01

New edge diagnostics and detailed analysis of magnetic topology have significantly improved the comprehension of the processes developing at the boundary of a reversed-field pinch (RFP) plasma in RFX-mod (a = 0.46 m, R = 2 m). An upper critical density n C ∼ 0.4 n G (n G Greenwald density) is found to limit the operational space for the improved quasi-single helical (QSH) regime: magnetic topology reconstructions and diagnostic observations suggest that this limit is due to a helical plasma-wall interaction which determines toroidally and poloidally localized edge density accumulation and cooling. The experimental evidence is provided by a variety of diagnostics: the magnetic boundary as reconstructed from equilibrium codes reveals a helical deformation, which is well correlated with the modulation of edge pressure profile as reconstructed from the thermal helium beam diagnostic. Correlations with the helical deformation are also observed on the space- and time-resolved patterns of the floating potential measured at the wall, and with the edge plasma flow, obtained from different diagnostics. The relevance of these findings is that understanding the mechanisms that limit the operational space of QSH is decisive in achieving the goal of high-density stationary helical RFP equilibrium.

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)

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

The Electronic Library of the Thermal Physical Databases

International Nuclear Information System (INIS)

Zhuravleva, Y.; Mingaleeva, G.; Mokrousov, K.; Yashnikov, D.

2008-01-01

Up-to-date quality assurance procedure requires the permanent verification of the best-estimate thermal-hydraulic system codes and the uncertainty analysis of results. Therefore, the researches need the growing up amount of the experimental data. Over the last years RDIPE has been carried out the verification of RELAP5/mod3.2 code and safety analysis for NPP with RBMK reactor. Moreover, these activities include both Russian (Puchok, Korsar, RATEG) and foreign codes (RELAP, MELCOR, ATHLET). Such activities require of the accumulation and the assessment of the large amount of experimental data. Electronic data base library was created in order to unify and keep the large amount of the primary experimental data. The special attention was given to completeness and sufficiency of information for modelling of the experiments. Generally this activity was carried out in the collaboration with the authors of experiment. First of all the experimental data for the additional verification of Russian and foreign codes relating to RBMK reactor safety analysis were included in the library. The following phenomena are specific and important: outflow from the main circulation circuit including critical flow of water, two phases mixture and vapour through the break, flow limiters, long channels with/ without local resistance and other circuit elements; thermal hydraulic process in reactor channels: pressure-drop, relative movement of phases, countercurrent flow, reflooding; heat transfer in fuel bundles including radiation heat transfer; heat transfer before and after critical heat flux transition in the rod bundle; variation of steam-water level in drum separator. These phenomena were studied at the test sites of KPI (Ukraine), Lithuanian Energy Institute, RDIPE (Russia), Russian Research Center 'Kurchatov Institute', EREC (Russia) and others. Transient modes data from operating power plants became the important part of the library. The authors of the electronic thermal physical

Predictable anomalies of process parameters on failure mode of internal structures in RPV by transient thermal-hydraulic analysis

International Nuclear Information System (INIS)

Maki, Akira; Mori, Michitsugu; Kanemoto, Shigeru; Konishi, Hideo

1997-01-01

A study has been conducted to evaluate how process parameters will exhibit the change in the event of the troubles related to reactor internal by using transient thermal-hydraulic analysis codes (RETRAN3D-MOD002, etc.). In the present study, the following six events are analytically investigated: 1) a leak from the upper plenum; 2) a leak from the middle part of a shroud; 3) a leak from the lower plenum; 4) a leak from the riser pipe for the jet-pump; 5) the blockage of the jet-pump nozzle; and 6) a leak from the jet-pump diffuser. The results by analyses indicated that the leak from the upper plenum resulted in increasing in the inlet temperature of primary loop recirculation (PLR) and in the differential pressure at the core support plate, and decreasing in the neutron flux (reactor power). Similar analyses were made for the five other events to identify the pattern of relevant process parameter variation in each event. (author)

An evapotranspiration product for arid regions based on the three-temperature model and thermal remote sensing

Science.gov (United States)

Xiong, Yu Jiu; Zhao, Shao Hua; Tian, Fei; Qiu, Guo Yu

2015-11-01

An accurate estimation of evapotranspiration (ET) is crucial to better understand the water budget and improve related studies. Satellite remote sensing provides an unprecedented opportunity to map the spatiotemporal distribution of ET. However, ET values from barren or sparsely vegetated areas in arid regions are often assumed to be zero in typical ET products because of their low values. In addition, separating ET into soil evaporation (Es) and vegetation transpiration (Ec) is difficult. To address these challenges, we developed an ET product (MOD3T) based on a three-temperature model and thermal remote sensing, specifically Moderate Resolution Imaging Spectroradiometer (MODIS) data. MOD3T has a spatial resolution of 1 km and a temporal resolution of 8 days. All input parameters except air temperature were obtained from MODIS datasets. Validation in two adjacent arid river basins in northwestern China showed that the mean absolute errors (mean absolute percent errors) between the MOD3T and flux tower ET were 0.71 mm d-1 (18.5%) and 0.16 mm d-1 (24.9%) for a densely vegetated area and sparsely vegetated sandy desert, respectively. The error between the MOD3T and water balance ET was 24 mm y-1 (8.1%). The Ec/ET or Es/ET of MOD3T was comparable to the observed stable oxygen and hydrogen isotopes. Unlike the MODIS ET (MOD16), MOD3T could not provide continuous ET values (as 70% of the MOD16 area lacked data) but exhibited relatively low uncertainty, particularly in cold seasons. Therefore, MOD3T can provide ET, Es and Ec estimates for arid regions within acceptable ranges.

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....

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.

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

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

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

Thermodynamical analysis of human thermal comfort

International Nuclear Information System (INIS)

Prek, Matjaz

2006-01-01

Traditional methods of human thermal comfort analysis are based on the first law of thermodynamics. These methods use an energy balance of the human body to determine heat transfer between the body and its environment. By contrast, the second law of thermodynamics introduces the useful concept of exergy. It enables the determination of the exergy consumption within the human body dependent on human and environmental factors. Human body exergy consumption varies with the combination of environmental (room) conditions. This process is related to human thermal comfort in connection with temperature, heat, and mass transfer. In this paper a thermodynamic analysis of human heat and mass transfer based on the 2nd law of thermodynamics in presented. It is shown that the human body's exergy consumption in relation to selected human parameters exhibits a minimal value at certain combinations of environmental parameters. The expected thermal sensation also shows that there is a correlation between exergy consumption and thermal sensation. Thus, our analysis represents an improvement in human thermal modelling and gives more information about the environmental impact on expected human thermal sensation

RELAP5 thermal-hydraulic analyses of overcooling sequences in a pressurized water reactor

International Nuclear Information System (INIS)

Bolander, M.A.; Fletcher, C.D.; Davis, C.B.; Kullberg, C.M.; Stitt, B.D.; Waterman, M.E.; Burtt, J.D.

1984-01-01

In support of the Pressurized Thermal Shock Integration Study, sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.6 and MOD2.0 computer codes. These analyses were performed for the H.B. Robinson Unit 2 pressurized water reactor, which is a Westinghouse 3-loop design plant. Results of the RELAP5 analyses are presented. The capabilities of the RELAP5 computer code as a tool for analyzing integral plant transients requiring a detailed plant model, including complex trip logic and major control systems, are examined

RESEARCH OF INFLUENCE OF THE HIGH-SPEED THERMAL PROCESSING REGIMES ON STRUCTURE AND MECHANICAL PROPERTIES OF PIPE STEEL 32G2

Directory of Open Access Journals (Sweden)

A. I. Gordienko

2012-01-01

Full Text Available Researches on influence of high-speed heating temperature, regimes of cooling and temperature of abatement on structure and mechanical properties of pipe steel 32G2 are carried out. Recommendations on the regimes of high-speed thermal processing of steel 32G2 which can be used at manufacturing of seamless pipes are given.

Fatigue resistance and crack propensity of large MOD composite resin restorations: direct versus CAD/CAM inlays.

Science.gov (United States)

Batalha-Silva, Silvana; de Andrada, Mauro Amaral Caldeira; Maia, Hamilton Pires; Magne, Pascal

2013-03-01

To assess the influence of material/technique selection (direct vs. CAD/CAM inlays) for large MOD composite adhesive restorations and its effect on the crack propensity and in vitro accelerated fatigue resistance. A standardized MOD slot-type tooth preparation was applied to 32 extracted maxillary molars (5mm depth and 5mm bucco-palatal width) including immediately sealed dentin for the inlay group. Fifteen teeth were restored with direct composite resin restoration (Miris2) and 17 teeth received milled inlays using Paradigm MZ100 block in the CEREC machine. All inlays were adhesively luted with a light curing composite resin (Filtek Z100). Enamel shrinkage-induced cracks were tracked with photography and transillumination. Cyclic isometric chewing (5 Hz) was simulated, starting with a load of 200 N (5000 cycles), followed by stages of 400, 600, 800, 1000, 1200 and 1400 N at a maximum of 30,000 cycles each. Samples were loaded until fracture or to a maximum of 185,000 cycles. Teeth restored with the direct technique fractured at an average load of 1213 N and two of them withstood all loading cycles (survival=13%); with inlays, the survival rate was 100%. Most failures with Miris2 occurred above the CEJ and were re-restorable (67%), but generated more shrinkage-induced cracks (47% of the specimen vs. 7% for inlays). CAD/CAM MZ100 inlays increased the accelerated fatigue resistance and decreased the crack propensity of large MOD restorations when compared to direct restorations. While both restorative techniques yielded excellent fatigue results at physiological masticatory loads, CAD/CAM inlays seem more indicated for high-load patients. Copyright © 2012 Academy of Dental Materials. Published by Elsevier Ltd. All rights reserved.

Thermal Analysis of TRIO-CINEMA Mission

Directory of Open Access Journals (Sweden)

Jaegun Yoo

2012-03-01

Full Text Available Thermal analysis and control design are prerequisite essential to design the satellite. In the space environment, it makes satellite survive from extreme hot and cold conditions. In recent years CubeSat mission is developed for many kinds of purpose. Triplet Ionospheric Observatory (TRIO–CubeSat for Ion, Neutral, Electron, MAgnetic fields (CINEMA is required to weigh less than 3 kg and operate on minimal 3 W power. In this paper we describe the thermal analysis and control design for TRIO-CINEMA mission. For this thermal analysis, we made a thermal model of the CubeSat with finite element method and NX6.0 TMG software is used to simulate this analysis model. Based on this result, passive thermal control method has been applied to thermal design of CINEMA. In order to get the better conduction between solar panel and chassis, we choose aluminum 6061-T6 for the material property of standoff. We can increase the average temperature of top and bottom solar panels from -70°C to -40°C and decrease the average temperature of the magnetometer from +93°C to -4°C using black paint on the surface of the chassis, inside of top & bottom solar panels, and magnetometer.

SCDAP/RELAP5 code development and assessment

International Nuclear Information System (INIS)

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

1996-01-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The current version of the code is SCDAP/RELAP5/MOD3.1e. Although MOD3.1e contains a number of significant improvements since the initial version of MOD3.1 was released, new models to treat the behavior of the fuel and cladding during reflood have had the most dramatic impact on the code's calculations. This paper provides a brief description of the new reflood models, presents highlights of the assessment of the current version of MOD3.1, and discusses future SCDAP/RELAP5/MOD3.2 model development activities

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...

Fluctuating zonal flows in the I-mode regime in Alcator C-Mod

Energy Technology Data Exchange (ETDEWEB)

Cziegler, I.; Diamond, P. H.; Fedorczak, N.; Manz, P.; Tynan, G. R.; Xu, M. [Center for Momentum Transport and Flow Organization, University of California, San Diego, La Jolla, California 92093 (United States); Churchill, R. M.; Hubbard, A. E.; Lipschultz, B.; Sierchio, J. M.; Terry, J. L.; Theiler, C. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02138 (United States)

2013-05-15

Velocity fields and density fluctuations of edge turbulence are studied in I-mode [F. Ryter et al., Plasma Phys. Controlled Fusion 40, 725 (1998)] plasmas of the Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] tokamak, which are characterized by a strong thermal transport barrier in the edge while providing little or no barrier to the transport of both bulk and impurity particles. Although previous work showed no clear geodesic-acoustic modes (GAM) on C-Mod, using a newly implemented, gas-puff-imaging based time-delay-estimate velocity inference algorithm, GAM are now shown to be ubiquitous in all I-mode discharges examined to date, with the time histories of the GAM and the I-mode specific [D. Whyte et al., Nucl. Fusion 50, 105005 (2010)] Weakly Coherent Mode (WCM, f = 100–300 kHz, Δf/f≈0.5, and k{sub θ}≈1.3 cm{sup −1}) closely following each other through the entire duration of the regime. Thus, the I-mode presents an example of a plasma state in which zero frequency zonal flows and GAM continuously coexist. Using two-field (density-velocity and radial-poloidal velocity) bispectral methods, the GAM are shown to be coupled to the WCM and to be responsible for its broad frequency structure. The effective nonlinear growth rate of the GAM is estimated, and its comparison to the collisional damping rate seems to suggest a new view on I-mode threshold physics.

RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1

International Nuclear Information System (INIS)

1995-08-01

The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes

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

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

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

Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

Energy Technology Data Exchange (ETDEWEB)

Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

2017-04-15

The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

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, ...

Thermal and Alignment Analysis of the Instrument-Level ATLAS Thermal Vacuum Test

Science.gov (United States)

Bradshaw, Heather

2012-01-01

This paper describes the thermal analysis and test design performed in preparation for the ATLAS thermal vacuum test. NASA's Advanced Topographic Laser Altimeter System (ATLAS) will be flown as the sole instrument aboard the Ice, Cloud, and land Elevation Satellite-2 (ICESat-2). It will be used to take measurements of topography and ice thickness for Arctic and Antarctic regions, providing crucial data used to predict future changes in worldwide sea levels. Due to the precise measurements ATLAS is taking, the laser altimeter has very tight pointing requirements. Therefore, the instrument is very sensitive to temperature-induced thermal distortions. For this reason, it is necessary to perform a Structural, Thermal, Optical Performance (STOP) analysis not only for flight, but also to ensure performance requirements can be operationally met during instrument-level thermal vacuum testing. This paper describes the thermal model created for the chamber setup, which was used to generate inputs for the environmental STOP analysis. This paper also presents the results of the STOP analysis, which indicate that the test predictions adequately replicate the thermal distortions predicted for flight. This is a new application of an existing process, as STOP analyses are generally performed to predict flight behavior only. Another novel aspect of this test is that it presents the opportunity to verify pointing results of a STOP model, which is not generally done. It is possible in this case, however, because the actual pointing will be measured using flight hardware during thermal vacuum testing and can be compared to STOP predictions.

Analysis of select Mod-1 semiscale blowdown heat transfer tests. Final report

International Nuclear Information System (INIS)

Irani, A.A.; Fujita, N.; Mecham, D.C.; Ching, J.T.; Gose, G.C.; Hentzen, R.D.; Sawtelle, G.R.; Moore, K.V.

1976-10-01

The report contains the RELAP4 analysis and sensitivity studies of Semiscale Tests S-02-2 and S-02-7. The Semiscale System is an electrically heated experiment designed to produce data on system performance typical of PWR thermal-hydraulic behavior. The RELAP4 program used for these analyses is a digital computer program developed to predict the thermal-hydraulic behavior of experimental systems and water-cooled nuclear reactors subjected to postulated transients. The results of the analysis for Test S-02-2 were in very good agreement with the data. Two parameters which required improvement were identified. These were the lower plenum density and the mass flow on the vessel side of the break. Subsequently, before analyzing Test S-02-7, the lower plenum was renodalized and the critical flow model at the vessel side break was modified. The results of the analysis of Test S-02-7 compared more favorably with the data than those of S-02-2. Additional sensitivity studies included time step studies, steam generator and downcomer modeling, and core nodalization

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.

Integrated numerical design of an innovative Lower Hybrid launcher for Alcator C-Mod

International Nuclear Information System (INIS)

Meneghini, O.; Shiraiwa, S.; Beck, W.; Irby, J.; Koert, P.; Parker, R. R.; Viera, R.; Wukitch, S.; Wilson, J.

2009-01-01

The new Alcator C-Mod LHCD system (LH2) is based on the concept of a four way splitter [1] which evenly splits the RF power among the four waveguides that compose one of the 16 columns of the LH grill. In this work several simulation tools have been used to study the LH2 coupling performance and the launched spectra when facing a plasma, numerically verifying the effectiveness of the four way splitter concept and further improving its design. The TOPLHA code has been used for modeling reflections at the antenna/plasma interface. TOPLHA results have been then coupled to the commercial code CST Microwave Studio to efficiently optimize the four way splitter geometry for several plasma scenarios. Subsequently, the COMSOL Multiphysics code has been used to self consistently take into account the electromagnetic-thermal-structural interactions. This comprehensive and predictive analysis has proven to be very valuable for understanding the behavior of the system when facing the plasma and has profoundly influenced several design choices of the LH2. According to the simulations, the final design ensures even poloidal power splitting for a wide range of plasma parameters, which ultimately results in an improvement of the wave coupling and an increased maximum operating power.

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

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

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

Thermal sensor properties of PANI (EB)–CSAX (X= 0⋅ 4±0⋅ 1 mol ...

Indian Academy of Sciences (India)

Home; Journals; Bulletin of Materials Science; Volume 25; Issue 6 ... superior to ceramic metal (Cermet) films, prepared by metallo organic deposition (MOD) ... these superior thermal-sensing properties together with optical studies and surface ...

Validation of a thermal-hydraulic system code on a simple example

International Nuclear Information System (INIS)

Kopecek, Vit; Zacha, Pavel

2014-01-01

A mathematical model of a U tube was set up and the analytical solution was calculated and used in the assessment of the numerical solutions obtained by using the RELAP5 mod3.3 and TRACE V5 thermal hydraulics codes. A good agreement between the 2 types of calculation was obtained.

32 CFR 989.32 - Noise.

Science.gov (United States)

2010-07-01

... 32 National Defense 6 2010-07-01 2010-07-01 false Noise. 989.32 Section 989.32 National Defense... ANALYSIS PROCESS (EIAP) § 989.32 Noise. Aircraft noise data files used for analysis during EIAP will be... System for Aircraft Noise for military training routes and military operating areas. Guidance on...

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...

Assessment of Human Bio-Behavior During Gait Process Using LifeMOD Software

Directory of Open Access Journals (Sweden)

Liliana Rogozea

2011-01-01

Full Text Available In this paper we present a set of observations concerning the
analysis and assessment of human bio-behavior during gait process. In the first part of the paper the fundamental and theoretical considerations of the gait process are approached and aspects connected to malfunctions are expressed. In the second part of the paper we present the modeling methodology using
the LifeMOD software, while in the third part the results and conclusions are presented.

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)

A review of analyses of LOFT and semiscale tests made at IDAHO National Engineering Laboratory using RELAP5/MOD1

International Nuclear Information System (INIS)

Hall, P.C.

1984-03-01

Within the LOFT and Semiscale programs at INEL, many post-test analysis calculations have been performed using RELAP5/MOD1. In this report, these calculations are reviewed from the standpoint of assessing the performance of the code. Because the calculations were spread over a number of years, different cycles of RELAP5/MOD1 have been employed. Rather than explicitly assessing several cycles of the code, a more general view has been adopted and an attempt has been made to identify those areas in which the code is systematically successful or alternatively, frequently experiences difficulties. (author)

Experiment data report for semiscale Mod-1 test S-04-1 (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-04-1 of the Semiscale Mod-1 Baseline ECC Test Series. This test is 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 system. Test S-04-1 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 system volume scaled coolant injection parameters. 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 in such a manner as to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

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

Modelling and analysis of a compensator burst after a check valve slam with the pressure surge code DYVRO mod. 3

International Nuclear Information System (INIS)

Neuhaus, Thorsten; Schaffrath, Andreas

2009-01-01

In this contribution the analysis and calculation of a compensator burst after a pump start and check valve slam with the pressure surge code DYVRO mod. 3 are presented. The compensator burst occurred in the essential service water system (ESWS) of a pressurized water reactor (PWR) in a deviant operation mode. Due to lack of knowledge about the causes a systematic investigation has been performed by TUV NORD SysTec GmbH and Co. KG. The following scenario was identified as most likely: Because of maintenance a heat exchanger was shut off from the ESWS by a closed valve. Due to the hydrostatic pressure profile air had been sucked in through this leaky closed valve forming an air bubble. After the pump start the water was accelerated against the closed valve where the air bubble was compressed. The subsequent backflow resulted in a fast closing of a check valve and a pressure surge that caused the compensator burst. Calculations have been performed with the self developed and validated pressure surge computer code DYVRO mod. 3. The present paper is focussed on the modelling of the pipe system, the pump, the check valve and the behaviour of the air bubble as well as the simulation of the incident. The calculated maximum pressure in the ESWS is above 3 MPa, which is approx. four times higher than the design pressure of 0.7 MPa. This pressure increase has led most likely to the abrupt compensator failure. (author)

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

International Nuclear Information System (INIS)

Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

2012-01-01

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

Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

International Nuclear Information System (INIS)

Fuller, R.; Harrell, J.

1996-01-01

The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves

Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

Energy Technology Data Exchange (ETDEWEB)

Fuller, R.; Harrell, J.

1996-12-01

The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

Foundation heat transfer analysis for buildings with thermal piles

International Nuclear Information System (INIS)

Almanza Huerta, Luis Enrique; Krarti, Moncef

2015-01-01

Highlights: • A numerical transient thermal model for thermo-active foundations is developed. • Thermal interactions between thermal piles and building foundations are evaluated. • A simplified analysis method of thermal interactions between thermal piles and building foundations is developed. - Abstract: Thermal piles or thermo-active foundations utilize heat exchangers embedded within foundation footings to heat and/or cool buildings. In this paper, the impact of thermal piles on building foundation heat transfer is investigated. In particular, a simplified analysis method is developed to estimate the annual ground-coupled foundation heat transfer when buildings are equipped with thermal piles. First, a numerical analysis of the thermal performance of thermo-active building foundations is developed and used to assess the interactions between thermal piles and slab-on-grade building foundations. The impact of various design parameters and operating conditions is evaluated including foundation pile depth, building slab width, foundation insulation configuration, and soil thermal properties. Based on the results of a series of parametric analyses, a simplified analysis method is presented to assess the impact of the thermal piles on the annual heat fluxes toward or from the building foundations. A comparative evaluation of the predictions of the simplified analysis method and those obtained from the detailed numerical analysis indicated good agreement with prediction accuracy lower than 5%. Moreover, it is found that thermal piles can affect annual building foundation heat loss/gain by up to 30% depending on foundation size and insulation level

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.

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.

Transient thermal analysis of Vega launcher structures

Energy Technology Data Exchange (ETDEWEB)

Gori, F. [University of Rome ' Tor Vergata' , Rome (Italy); De Stefanis, M. [Thales Alenia Space Italia, Rome (Italy); Worek, W.M. [University of Illinois at Chicago, Chicago (United States)], E-mail: wworek@uic.edu; Minkowycz, W.J. [University of Illinois at Chicago, Chicago (United States)

2008-12-15

A transient thermal analysis is carried out to verify the base cover thermal protection system of Vega 2nd stage Solid Rocket Motor (SRM) and the flange coupling of the inter-stage 2/3. The analysis is performed with a finite element code. The work has developed suitable numerical Fortran subroutines to assign radiation and convection boundary conditions. The thermal behaviour of the structures is presented.

Comparison and Evaluation of Annual NDVI Time Series in China Derived from the NOAA AVHRR LTDR and Terra MODIS MOD13C1 Products.

Science.gov (United States)

Guo, Xiaoyi; Zhang, Hongyan; Wu, Zhengfang; Zhao, Jianjun; Zhang, Zhengxiang

2017-06-06

Time series of Normalized Difference Vegetation Index (NDVI) derived from multiple satellite sensors are crucial data to study vegetation dynamics. The Land Long Term Data Record Version 4 (LTDR V4) NDVI dataset was recently released at a 0.05 × 0.05° spatial resolution and daily temporal resolution. In this study, annual NDVI time series that are composited by the LTDR V4 and Moderate Resolution Imaging Spectroradiometer (MODIS) NDVI datasets (MOD13C1) are compared and evaluated for the period from 2001 to 2014 in China. The spatial patterns of the NDVI generally match between the LTDR V4 and MOD13C1 datasets. The transitional zone between high and low NDVI values generally matches the boundary of semi-arid and sub-humid regions. A significant and high coefficient of determination is found between the two datasets according to a pixel-based correlation analysis. The spatially averaged NDVI of LTDR V4 is characterized by a much weaker positive regression slope relative to that of the spatially averaged NDVI of the MOD13C1 dataset because of changes in NOAA AVHRR sensors between 2005 and 2006. The measured NDVI values of LTDR V4 were always higher than that of MOD13C1 in western China due to the relatively lower atmospheric water vapor content in western China, and opposite observation appeared in eastern China. In total, 18.54% of the LTDR V4 NDVI pixels exhibit significant trends, whereas 35.79% of the MOD13C1 NDVI pixels show significant trends. Good agreement is observed between the significant trends of the two datasets in the Northeast Plain, Bohai Economic Rim, Loess Plateau, and Yangtze River Delta. By contrast, the datasets contrasted in northwestern desert regions and southern China. A trend analysis of the regression slope values according to the vegetation type shows good agreement between the LTDR V4 and MOD13C1 datasets. This study demonstrates the spatial and temporal consistencies and discrepancies between the AVHRR LTDR and MODIS MOD13C1 NDVI

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

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)

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

Edge Minority Heating Experiment in Alcator C-Mod

International Nuclear Information System (INIS)

Zweben, S.J.; Terry, J.L.; Bonoli, P.; Budny, R.; Chang, C.S.; Fiore, C.; Schilling, G.; Wukitch, S.; Hughes, J.; Lin, Y.; Perkins, R.; Porkolab, M.; Alcator C-Mod Team

2005-01-01

An attempt was made to control global plasma confinement in the Alcator C-Mod tokamak by applying ion cyclotron resonance heating (ICRH) power to the plasma edge in order to deliberately create a minority ion tail loss. In theory, an edge fast ion loss could modify the edge electric field and so stabilize the edge turbulence, which might then reduce the H-mode power threshold or improve the H-mode barrier. However, the experimental result was that edge minority heating resulted in no improvement in the edge plasma parameters or global stored energy, at least at power levels of P RF (le) 5.5 MW. A preliminary analysis of these results is presented and some ideas for improvement are discussed

Thermal Analysis of Solar Panels

Science.gov (United States)

Barth, Nicolas; de Correia, João Pedro Magalhães; Ahzi, Saïd; Khaleel, Mohammad Ahmed

In this work, we propose to analyze the thermal behavior of PV panels using finite element simulations (FEM). We applied this analysis to compute the temperature distribution in a PV panel BP 350 subjected to different atmospheric conditions. This analysis takes into account existing formulations in the literature and, based on NOCT conditions, meteorological data was used to validate our approach for different wind speed and solar irradiance. The electrical performance of the PV panel was also studied. The proposed 2D FEM analysis is applied to different region's climates and was also used to consider the role of thermal inertia on the optimization of the PV device efficiency.

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

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

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

Accident analysis in the water loop of the nuclear engineering department of IPEN using the RELAP4 code

International Nuclear Information System (INIS)

Fernandes Filho, T.L.

1980-06-01

A thermal-hydraulic analysis to describe the transient behavior in the water loop of the Nuclear Engineering Department of the Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo, Brazil, was performed. Postulated accidents such as those resulting from (1) loss of coolant, (2) main pump failure and (3) power excursions, were studied. The computer code RELAP4/Mod.3 was employed as the principal tool of analysis. (Author) [pt

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

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”.

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

International Nuclear Information System (INIS)

Analytis, G.Th.

2004-01-01

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

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.

”Giv vold mod kvinder det røde kort” - En kampagneanalyse

OpenAIRE

Dührr, Tine; Jeberg Kjær, Cecilie; Lassen Jørgensen, Katrine; Linea Pohl, Dina

2013-01-01

This project investigates if the campaign “Giv vold mod kvinder det røde kort” is successful in reaching its target audience. We keep a strict focus on the campaign’s Facebook page, because of its obvious role as the primary place where the target audience will encounter the campaign. We examine the Facebook page through a discourse analysis, to get a clear sense of the difficulties with reaching the target audience. Afterwards we examine the target audience, with a focus group interview to g...

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

Analysis of 4-strap ICRF Antenna Performance in Alcator C-Mod

International Nuclear Information System (INIS)

Schilling, G.; Wukitch, S.J.; Boivin, R.L.; Goetz, J.A.; Hosea, J.C.; Irby, J.H.; Lin, Y.; Parisot, A.; Porkolab, M.; Wilson, J.R.

2003-01-01

A 4-strap ICRF antenna was designed and fabricated for plasma heating and current drive in the Alcator C-Mod tokamak. Initial upgrades were carried out in 2000 and 2001, which eliminated surface arcing between the metallic protection tiles and reduced plasma-wall interactions at the antenna front surface. A boron nitride septum was added at the antenna midplane to intersect electric fields resulting from radio-frequency sheath rectification, which eliminated antenna corner heating at high power levels. The current feeds to the radiating straps were reoriented from an E||B to E parallel B geometry, avoiding the empirically observed ∼15 kV/cm field limit and raising antenna voltage holding capability. Further modifications were carried out in 2002 and 2003. These included changes to the antenna current strap, the boron nitride tile mounting geometry, and shielding the BN-metal interface from the plasma. The antenna heating efficiency, power, and voltage characteristics under these various configurations will be presented

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 ...

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

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

éomatériaux visqueux sous chargement monotone (fig. 3. 30, développé par l'École Centrale de Paris puis validé et utilisé par I'IFP pour des calculs de fondations. Il comprend un mécanisme de plastification isotrope basé sur le modèle Cam-Clay et deux variables d'écrouissage (comme les modèles précédents. Son originalité principale est la possibilité de décrire le fluage tertiaire grâce à une variable d'endommagement de la viscosité dont l'évolution est irréversible. Le modèle utilise 18 paramètres qui doivent être déterminés selon une procédure d'identification rigoureuse, à partir d'essais de laboratoire courants. Evalué sur chemins homogènes de compression à l'appareil triaxial et pour deux matériaux (fig. 3. 31, il donne globalement satisfaction. Il a aussi été utilisé avec succès pour analyser quelques aspects du comportements des sols gelés (fig. 3. 32. Le modèle VISCYC est une généralisation viscoplastique de CYCLADE qui a fourni de bons résultats sur des chemins cycliques non-drainés à l'appareil triaxial pour deux matériaux argileux (fi. 3. 33 à 3. 35. Ces modèles, une fois intégrés dans un programme utilisant la méthode des éléments finis, peuvent être utilisés pour le calcul du comportement des fondations marines sous chargement extrêmes ou de fatigue. Leur performance dans ce domaine est étudiée en détail dans la deuxième partie de cette synthèse qui fait l'objet d'un autre article.

Thermal analysis studies of ammonium uranyl carbonate

International Nuclear Information System (INIS)

Cao Xinsheng; Ma Xuezhong; Wang Fapin; Liu Naixin; Ji Changhong

1988-01-01

The simultaneous thermogravimetry and differential thermal analysis of the ammonium uranyl carbonate powder were performed with heat balance in the following atmosphers: Air, Ar and Ar-8%H 2 . The thermogravimetry and differential thermal analysis curves of the ammonium uranyl carbonate powder obtained from different source were reported and discussed

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)