m002 mod cmsx: Topics by WorldWideScience.org

Atom-probe field-ion microscopy investigation of CMSX-4 Ni-base superalloy laser beam welds

An investigation of the solidification behavior was carried out on the directionally solidified single crystal superalloy CMSX 6. The relationship between structure morphology and the process parameters has been experimentally determined and illustrated in a solidification diagram. The metallographic analyses of transverse sections within the solidification interval yield the sequence of phase formation and the evolution of solid fraction. The solidification process and the corresponding structure have been discussed in detail. (orig.)

1995-11-01

Repairing directionally solidified and single-crystalline gas turbine blades; Reparatur von gerichtet erstarrten und einkristallinen Gasturbinenschaufeln

CMSX-4 superalloy laser beam welds were investigated by transmission electron microscopy and atom probe field-ion microscopy (APFIM). The weld microstructure consisted of fine (10- to 50-nm) irregularly shaped {gamma}` precipitates (0.65 to 0.75 volume fraction) within the {gamma} matrix. APFIM compositions of the {gamma} and {gamma}` phases were found to be different from those in the base metal. Concentration profiles across the {gamma} and {gamma}` phases showed extensive variations of Cr, Co and Al concentrations as a function of distance within the {gamma} phase. Calculated lattice misfits near the {gamma}/{gamma}` interface in the welds are positive values compared to the negative values for base metal. (orig.).

1996-09-01

Assessment of RELAP5/MOD2 against a natural circulation experiment in Nuclear Power Plant Borssele. International Agreement Report

Investigations show that nickel-based alloys (MAR M 247 CC, CM DS 247 L and CMSX-4) can be successfully brazened. Tensile tests of MAR M 247 CC samples brazened with foil brazing showed values comparable to those of the base material, some samples failed because of base material problems. Tests with powder mixes made up from commercial brazens and the base material CMSX-4 showed good wetting and a perfect microstructure similar to the base material. Tensile test values at RT (test temperature) and 850 degrees centigrade show pronounced scatter. Tensile tests at RT and 850 degrees centigrade are used to optimse brazing cycles for the types of brazen used (temperature, time). (orig.) [German] Die Untersuchungen haben gezeigt, dass es moeglich ist, Nickelbasislegierungen (MAR M 247 CC, CM DS 247 LC und CMSX-4) erfolgreich zu loeten. Die im Zugversuch erhaltenen Festigkeitswerte der mit folienfoermigen Lot geloeteten MAR M 247 ...

1998-07-01

Turbine blades from monocrystalline materials. Part-project: Thermomechanical and isothermal fatigue of the monocrystalline gamma-hardened nickel base superalloys CMSX-6 and SC 16. Final report; Turbinenschaufeln aus Einkristallwerkstoffen. Teilprojekt: Thermomechanisches und isothermes Ermuedungsverhalten der einkristallinen {gamma}`-gehaerteten Nickelbasis-Superlegierungen CMSX-6 und SC16 (Stichversuche). Abschlussbericht

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.

1993-07-01

Technical Lio - NASA Technical Reports Server

Nickel base superalloys for turbine construction are exposed to maximum thermal and mechanical stresses. owing to a change of materials by one of the project partners, the project only comprised analyses of the alloy CMSX-6. Isothermal tensile and fatigue tests provided basic data on the alloy. Thermomechanical fatigue tests showed that material life and microstructural changes are clearly, and to an extreme extent, dependent on the type of combination of the two stress factors `temperature interval` and `total strain amplitude`, while the damage to the material tends to be a function of the maximum stresses for a given stress path. On the one hand, these results make it possible to obtain information real stresses from the structures of real blades. On the other hand, constructional measures can be taken in order to better adapt the stresses to which alloys are exposed to their characteristic potential. (orig./AKF) [Deutsch] Die im Turbinenbau eingesetzten ...

1997-12-31

20 Watt-Hour Per Pound Regenerative Fuel Cell

Mod IE did not affect the valve flow characteristics. C. BIPROPELLANT -VALVE EVALUATION. 1. Ball and Shaft Seal Evaluation ...

Low cycle fatigue behaviour of a 4th generation Ni-base single-crystal superalloy TMS-138

... for evaluation of the electrochemiral performance of the materials and components used in EOS Rechargeable Fuel Cell Model RHO-24AH-Mod ...

1972-03-01

Input deuteron states in Mo even isotopes

Low cycle fatigue (LCF) of a fourth-generation single-crystal (SC) Ni-base superalloy TMS-138 was studied by comparison with a typical third-generation (TMS-75) and a second-generation (CMSX-4) Ni-base SC superalloys. TMS-138 exhibits improved LCF behaviour under a condition of the R ratio of 0 at temperatures of 1073 K and 1173 K. The addition of refractory elements resulted in a remarkable improvement of the LCF properties compared to those of the reference superalloys due to the different microstructure developed in TMS-138. (orig.)

2004-07-01

Roles of bumpy field on collisionless particle confinement in helical-axis heliotrons

An attempt is taken to explain anomalies in "9"2Mo(d, n)"9"3Tc, "9"2Mo(d, #alpha#)"9"0Nb, "9"4Mo(d, n)"9"5Tc, "9"8Mo(d, n)"9"9Tc, "9"8Mo(d, p)"9"9Mo, "9"8Mo(d, #alpha#)"9"6Nb, "1"0"0Mo(d, p)"1"0"1Mo and "1"0"0Mo(d, n)"1"0"1Tc reactions with input states having a one-particle nature. Thin films saturated with molybdenum isotopes at the approximately 1 mgxcm"-"2 surface density are used as targets. The targets are irradiated by the extracted cyclotron beam. The deuteron energy is 5-12 MeV. The reaction cross sections are determined by the activation analysis method. Quasi-stationary levels of the nucleus-deuteron system are calculated. Weak anomalies revealing in a smooth (d, #alpha#) reaction cross section on sup(92, 98)Mo nuclei, which do not necessarily correlate with anomalies in the (d, n) and (d, p) channels, are ...

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

Roles of bumpy field on collisionless particle confinement in helical-axis heliotrons are investigated with the model magnetic field and particle orbit calculations in the Boozer coordinates. The mod-B{sub min} contours can be shifted in the major radius direction with the control of the bumpy field, where B{sub min} is the minimum value of |B| in the toroidal direction within one field period. The area of closed mod-B{sub min} contours is a useful measure to evaluate global collisionless particle confinement as long as the mod-B{sub min} contours connect toroidally. Negative value of ratio between the bumpy and the helicity components contributes to obtain the largest area of closed mod-B{sub min} contours for finite ratio between the toroidicity and the helicity components. The radial variation of the bumpy field attributes to realize a toroidally localized mod-B{sub min} ...

1999-02-01

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

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 ...

1993-12-31

Evaluation of validity of the RELAP5/MOD3 flow regime map for horizontal tubes

1992-09-29

Assessment of RELAP5/MOD2 against natural circulation experiments performed with the REWET-III facility

RELAP5/MOD3 code was developed for western type power water reactors with vertical steam generators. Thus, this code should be validated also for VVER design with horizontal steam generators. The validation work, which has been started in Lappeenranta University of Technology (LUT), has already shown some weaknesses of the code. For example the flow inside a steam generator horizontal tube in some accident cases is not correctly modelled by the code. It may be the result of erroneous prediction of the flow regime. The aim of the study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal tubes. (18 refs.).

1996-12-31

Assessment of RELAP5/MOD2 against natural circulation experiments performed with the REWET-III facility

Natural circulation experiments carried out in the REWET-III facility in 1985 have been used for RELAP5/MOD2 assessment. The REWET-III facility is a scaled-down model of VVER-440 type reactors. The facility consists of a pressure vessel in which the downcomer is simulated with an external pipe assembly, hot and cold legs with loop seals and a horizontal steam generator. The volume scaling factor compared to the reference reactor is 1:2333. The present paper summarizes the experiences gained in the RELAP5/MOD2 calculations of selected REWET-III single- and two-phase natural circulation experiments. The code's ability to represent the main phenomena of experiments in both cases was satisfactory.

1992-04-01

Assessment of RELAP5/MOD2 against natural circulation experiments performed with the REWET-III facility

Natural circulation experiments carried out in the REWET-III facility in 1985 have been used for RELAP5/MOD2 assessment. The REWET-III facility is a scaled-down model of VVER-440 type reactors. The facility consists of a pressure vessel in which the downcomer is simulated with an external pipe assembly, hot and cold legs with loop seals and a horizontal steam generator. The volume scaling factor compared to the reference reactor is 1:2333. The present paper summarizes the experiences gained in the RELAP5/MOD2 calculations of selected REWET-III single- and two-phase natural circulation experiments. The code`s ability to represent the main phenomena of experiments in both cases was satisfactory.

1992-04-01

1992-01-01

UK CDHF for STP UK

A statistical framework for modeling gene expression using chromatin features and application to modENCODE datasets

Original holdings are now kept by the World Data Centre C1 at RAL: http://wdcc1. bnsc.rl.ac.uk MODS DOY = 246 ADID_REF Not applicable LOGICAL_FILE_ID ...

UK PubMed Central (United Kingdom)

We develop a statistical framework to study the relationship between chromatin features and gene expression. This can be used to predict gene expression of protein coding genes, as well as microRNAs....Full Text Available

2011-01-01

Comparison of modelling the PMK-2 steam generator by codes RELAP and MELCOR

Analysis of the noncondensing gas effect on the heat transfer in a horizontal steam generator by means of the RELAP5/MOD3.2 code

The RELAP5/MOD2 and MELCOR 1.8.1 codes have been used for simulate the PMK total loss of feedwater with secondary bleed and feed experiments done in a scale-model WWER-440 test facility. Nodalization studies and studies on several-core modelling options were also done. Good agreement was found between the calculations done by RELAP5/MOD2 and MELCOR 1.8.1 JY codes. (orig.) (5 refs., 31 figs.).

1992-09-29

Large turbine blades in a new technology for environment-friendly gas turbines. Project: Behaviour of defects and small cracks. Final report GT1; Grosse Turbinenschaufeln neuer Technologie fuer umweltschonende Gasturbinen. Teilprojekt: Verhalten von Fehlstellen und kurzen Rissen. Abschlussbericht GT1

When analyzing the loss-of-coolant accidents at VVER reactor NPP the problem of the effect of noncondensable gases on heat transfer in a horizontal steam generator (HSG) is gaining in importance. Based on the RELAP5/MOD3.2 computer code one analyzed the experiments to condense steam-and-gas mixture in a HSG. The calculations are shown to predict satisfactorily duration of steam generator poisoning from noncondensable gas

2005-03-01

1 2 3 4

A 5{sup th} generation Ni-base single crystal superalloy with superior elevated temperature properties

As a partner in this project, Siemens Power Generation investigated the crack propagation characteristics of short cracks in CMSX4. This involved crack propagation tests under dynamic and static load, creep tests on samples of different geometries, and life calculations, with the following results: No typical short cracking behaviour was observed under dynamic loads. Short crack propagation at 950 C under static load is well described by the fracture mechanical parameter C* integral. Notched bar creep tests showed longer times to cracking for notched bars. [German] In dem Verbundprojekt hat Siemens Power Generation u.a. die Untersuchungen von Rissausbreitungsverhalten von Kurzenrissen in CMSX4 als Hauptaufgabe bekommen. Dabei wurden Rissausbreitungsversuche unter dynamischer und statischer Belastung, Kriechversuche an Proben verschiedener Geometrie und Lebensdauerberechnung durchgefuehrt. Die wesentlichen Ergenisse sind wie folgt kurz ...

2001-07-01

Ultra-high temperature strength properties on Mod.9Cr-1Mo steel

The National Institute of Materials Science (NIMS) has utilized the in-house alloy design program to develop a 5{sup th} generation Ni-base single crystal superalloy, TMS-196 with superior high temperature creep, thermo mechanical fatigue (TMF) and oxidation resistance by incorporating further ruthenium (Ru) and chromium (Cr) content over the compositions of 4{sup th} generation superalloys. With Ru additions in advanced superalloys to enhance phase stability, higher content of refractory elements can be accommodated to provide further strengthening; the associated oxidation resistance can be improved by the increase in Cr additions. In present article, TMS-196 has been subjected to cyclic/isothermal oxidation tests at 1100 C and 900 C, creep at conditions between 800 C{proportional_to}1100 C/137MPa{proportional_to}735MPa and TMF cycles. Preliminary studies indicate that the surface of oxidized TMS-196 can exhibit continuous alumina layer attributing good oxidation resistance similar ...

2006-07-01

An Assessment of Through Thickness Mechanical Properties in Forged Thick Section Mod. 9Cr-1Mo Steel

A sodium-water reaction drove from the single tube break in steam generator of FBR might overheat labor tubes rapidly under internal pressure loadings. If the temperature of tube wall becomes too high, it has to be evaluated that the stress of tube does not exceed the material strength limit to prevent the propagation of tube rupture. This study clarified the tensile and creep properties of Mod.9Cr-1Mo steel at ultra-high temperature which will be used in evaluation of the tube burst by sodium-water reaction. The strain rates for tensile test are from 10%/min to 10%/sec, and creep-rupture time is maximum 277sec. The range of test temperature is 700degC to 1300degC. The main results obtained were as follows; (1) The evaluation data on the relationship between tensile strength and strain rate and creep-rupture strength in shorter time on Mod.9Cr-1Mo steel were acquired. (2) Short-term mechanical properties of Mod.9Cr-1Mo ...

2000-03-01

Study on tritium activity build-up in moderator and primary heat transport systems in 540 MWe reactor

Ferritic/martensitic steel, modified 9Cr-1Mo steels have been used most extensively in the power generation industry throughout the world due to having superior high temperature properties such as high strength, creep resistance, and good microstructure stability. These steels are also the primary candidate for the RPVs(Reactor Pressure Vessels) of High Temperature Gas-Cooled Reactors. Currently, many studies has been conducted in laboratory-scale for mod. 9Cr-1Mo steels. However, there is a lack of the study on forged thick- section for RPVs. The differences in characteristics including the through thickness microstructure and mechanical properties between internal and external locations may occur during cooling after austenitization, because the thickness of RPVs is over about 200mm. Therefore, in order to use ferritic/martensitic steel as RPVs, a detailed assessment of the through thickness properties is needed. The purpose of this study is to investigate the ...

2010-10-01

Study on tritium activity build-up in moderator and primary heat transport (PHT) systems in 540 MWe reactor

Tarapur Atomic Power Station unit-4 is first 540 MWe pressurized heavy water reactor. Heavy water is used as the coolant and moderator. With reactor operation tritium is formed by absorption of neutron by Deuterium atoms. Experience in the 220 MWe indicates that tritium is main contributor to the internal dose of radiation workers. Study on Tritium build up in Primary Heat Transport (PHT) and Moderator (MOD) system was carried out at the initial stage of the operation of the unit-3 and 4. This paper brings out tritium activity buildup in the PHT and MOD systems and its comparison with 220 MWe reactors. This paper helps in estimation of the internal dose contribution to the radiation workers at TAPS 3 and 4. (author)

2006-11-13

RELAP5/MOD3.1 and APROS 3.0 analyses of SBLOCA in scaled VVER-440 geometry

Tarapur Atomic Power Station unit-4 is first 540 MWe pressurized heavy water reactor. Heavy water is used as the coolant and moderator. With reactor operation tritium is formed by absorption of neutron by deuterium atoms. Experience in the 220 MWe indicates that tritium is main contributor to the internal dose of radiation workers. Study on tritium build up in Primary Heat Transport (PHT) and Moderator (MOD) system was carried out at the initial stage of the operation of the unit-4. This paper brings out tritium activity buildup in the PHT and MOD systems and its comparison with 220 MWe reactors. This paper helps in estimation of the internal dose contribution to the radiation workers at TAPS 3 and 4. (author)

2005-11-23

RELAP5/MOD3.1 and APROS 3.0 analyses of SBLOCA in scaled VVER-440 geometry

A cold-leg small-break loss-of-coolant accident (SBLOCA) experiment was performed on the PACTEL facility to study the behavior of natural circulation in a VVER-440 reactor geometry. The facility is a volumetrically scaled (1:305) integral test loop simulating the VVER-440 reactors used in Finland. The test results were used to assess the computer codes RELAP5/MOD3.1 and APROS 3.0 for VVER reactors. The behavior of the horizontal steam generator and the effect of the hot-leg loop seal were of particular interest. The specific parameters to be compared included the primary pressure and the downcomer mass flow rate.

1995-12-31

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

1995-11-01

Calculation model testing for the case of rcs hot collector rupture inside the horizontal steam generator of VVER-440 NPP

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

1994-06-01

Calculation model testing for the case of rcs hot collector rupture inside the horizontal steam generator of VVER-440 NPP

The calculations presented are based on RELAP5/MOD2-3 input for VVER 440/213 Bohunice NPP, developed within the framework of IAEA TC Project by an international team of specialists from CSFR, Hungary, Bulgaria and Poland. Project activities were condentrated on input data refinement and testing. Several cases were calculated using the latest version of RELAP5/MOD2 provided by RMA, Albuquerque to investigate some modelling assumptions, such as break location, geometrical representation of secondary circuit piping as well as the effect of deactivation of the signal controlling the SG isolation valves. (2 refs., 21 figs., 2 tabs.).

1993-12-31

British Nuclear Fuels PLC: report and accounts 1988-89

1992-09-29

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

This item covers a meeting held between members of the United Kingdom government's energy committee and representatives of British Nuclear Fuels (BNFL) to discuss their Annual Report and Accounts for the year 1988-89. The committee explored the reasons for escalating predictions of the costs of nuclear power and why decommissioning costs are so difficult to estimate accurately so as to include them in cost per kilowatt hour of generated electricity. The relationship between BNFL and the Ministry of Defence (MoD) was explored, as was the MoD's relationship with the United States Department of Defense. BNFL's financial position should improve when the thermal oxide reprocessing plant at Sellafield becomes operational, and the Chapelcross and Calder Hall reactors may contribute income from electricity generation. (UK).

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

Three experiments were conducted to study the behaviour of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes so that the results may be compared. Only the steam generator was modeled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments. (orig.) 4 refs.

1997-12-01

Nodalization schemes for PGV-440 steam generator model with RELAP5/MOD3

1995-09-10

Nodalization schemes for PGV-440 steam generator model with RELAP5/MOD3

Results of calculation of steady thermal-hydraulic characteristics of PVG-440 horizontal steam generator are presented. Steam flows in selected sections are compared to data provided by OKB Gidropress Calculated vapor void fractions are compared to measured ones. (orig.) (3 refs., 3 figs., 8 tabs.).

1993-12-31

Modelling the horizontal steam generator with APROS simulation code

1992-09-29

Modelling the horizontal steam generator with APROS simulation code

The Finnish simulation code APROS and especially its 5-equation model is applied to modelling the horizontal steam generator. Different nodalizations are used in the secondary side of different models. Simulation results of the stationary state run are compared with results of RELAP5/MOD2 calculations and with an experimental plant data. (2 refs., 3 figs., 4 tabs.).

1993-12-31

Management considerations of the large primary-to-secondary leakage accidents

1992-09-29

EOP-MOD Ver 2.0 2444055.5 1.0 11875 UT1-TAI UNDEF # Earth ...

The management procedure of a large PRISE (Primary-to-Secondary) leakage accident at Loviisa nuclear power plant taking into account the plant modifications which are expected to be realized during 1995-96 is described. The management procedure has been validated by performing thermal hydraulic analyses with the computer code RELAP5/MOD3 and the results from these analyses are also shortly discussed. (4 refs., 6 figs., 1 tab.).

1993-12-31

RELAP5/MOD3 code manual. Volume 4, Models and correlations

... 2.3385 -27064396 0.0037 0.0030 15. .000 .000 .000 2449038.5 1.9466 2.3213 - 27066977 0.0037 0.0035 12. .000 .000 .000 2449039.5 1.9312 2.3064 -27069688 ...

1 2 3 4

Performance of Alcator C-Mod core Thomson scattering system

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 presents modeling theory and associated numerical schemes; Volume II details instructions for code application and input data preparation; Volume III presents the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume IV ...

1995-08-05

Microscopic investigations of microcrack formation in two ferritic-martensitic steels in nonirradiated and irradiated conditions; Mikroskopische Untersuchung der Bildung von Ermuedungsrissen an zwei ferritisch-martensitischen Staehlen im unbestrahlten und vorbestrahlten Zustand

Design of the Alcator C-Mod Thomson scattering (TS) diagnostic is discussed and the results of the measurements are presented. The TS system has six spatial channels with observation volumes evenly distributed between the midplane and the edge of the plasma. Each channel is capable of measuring the electron density in the range N{sub e}=5{times}10{sup 19}{endash}5{times}10{sup 21} m{sup {minus}3} and temperature from T{sub e}=200 eV to 10 keV. A 30 Hz, 1.5 J per pulse Nd-YAG laser is employed allowing the measurements of evolution of T{sub e} and N{sub e} profiles during plasma shot. A laser beam position control and feedback system provides for the beam alignment stability and reliable electron density measurements. Examples of the core density and temperature profiles measured at different stages of the plasma evolution are discussed. {copyright} {ital 1999 American Institute of Physics.}

1999-01-01

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

The low-cycle fatigue properties of the ferritic-martensitic steels MANET I and II and F82H-mod, have been investigated at a temperature of 200 C. The strain-controlled fatigue has been performed with strain ranges between 0,4% and 1,0% and at a strain rate of 8.10{sup -4} s{sup -1}. The steel F82H-mod, is characterised by a higher plastic strain amplitude compared to MANET I and II. Some of the specimens have beenirradiated with a 104 MeV {alpha}-particle beam. The homogeneous implantation of 400 appm He leads to a hardening of the material. At the beginning of the fatigue tests the stress response of the irradiatedspecimens is higher compared to the unirradiated specimens. In general, thematerial shows softening during fatigue. The softening of irradiated material, which is fatigue tested with a high total strain amplitude, is higher than that of the unirradiated material. The lifetime is reduced. For irradiated specimens fatigued at small ...

1997-09-01

Comparisons of two-phase microgravity calculations using current and new flow regime maps in RELAP5

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 predicted 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 applied for this particular application includes heat transfer ...

2001-03-01

Comparisons of two-phase microgravity calculations using current and new flow regime maps in RELAP5

Recently, many experiments designed to quantify the parameters involved in microgravity two-phase flow have been performed. These experiments give significant insight to the differences between the flow regimes in 1-g and microgravity. However, the new correlations for pressure drop, heat transfer, and the flow regime maps are yet to be implemented into analytical methods. The purpose of this study is to model a KC-135 microgravity experiment, using the thermal-hydraulic does RELAP5/MOD2 and ATHENA. A comparison of these experimental results to code calculations from RELAP5/MOD2 and ATHENA is shown. Results show little difference between the ATHENA and the RELAP5 calculations. Also, modifications are made to the two-phase flow regime map in RELAP5 to model microgravity predictions. There is a substantial difference between the code's calculation before and after the changes were implemented. The heat transfer correlations of the code ...

1988-01-01

BWNT assessment of TRAC/PF1-MOD2

1988-11-04

Assessment of RELAP5/MOD3/CANDU"+ to Wolsung-1 D_2O leakage event

The TRAC/PFI-MOD2 Version 5.3 code was assessed against six FLECHT-SEASET forced reflood tests (31504, 31203, 31302, 31701, 34209, and 31922) and two cylindrical core test facility (CCTF) tests [C1-19 and C2-6]. The objective of this study was to evaluate the clad thermal response predictive capabilities of the code with the newly added reflood model under large-break loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (PWR). The TRAC model for the FLECHT-SEASET test facility was developed from a RELAP5 model. The test section was modeled using a vessel component with 23 axial levels, 1 radial ring, and 1 azimuthal cell. Test inlet and exit conditions were modeled using fill and break components, respectively. The measured lower and upper plenum test conditions were input to the model. The electrically heated rod was modeled using a rod component with 22 axial mesh points. The axial boundary of each mesh point coincided with a fluid cell ...

1993-11-14

High temperature strength properties of Mod.9Cr-1Mo steel forging weld zone

In order to evaluate the integrated performance of RELAP5/MOD3/CANDU"+ for CANDU operational transient analysis, we assesed the code to the D_2O leakage event occurred at Wolsung-I, 600 MW(e) CANDU reactor, on Oct. 20, '94. D_2O leakage event was initiated by stuck opening of liquid relief valve No.4 in primary coolant pressure and level control system. Assessment calculation was performed for the plant transients up to 1000 seconds after the initiating event. Calculation results are compared with those measured in primary heat transport system, pressure and inventory control system and boiler secondary system. Comparison with the plant trip log shows that the RELAP5/CANDU"+ is able to simulate the plant transients properly, from which we can conclude that the RELAP5/CANDU"+ is validated for application to CANDU operational transient analysis. CANDU specific models used in the assessment are fuel bundle heat transfer model, decay heat model and MOV(Motor Operated ...

2001-10-01

Effective thickness of CeO{sub 2} buffer layer for YBCO coated conductor by advanced TFA-MOD process

In the feasibility study on commercialized FBR systems, the application of 12Cr steel, with its physical properties and high-temperature strength properties, as structural material is considered in order to greatly reduce construction costs by compacted instruments and structures. In this report, creep, fatigue and the creep-fatigue properties of Mod.9Cr-1Mo steel forging (thickness 550mm) weld, representing conventional steels, were evaluated in order to obtain the needed basic data for the evaluation of the high-temperature strength of 12Cr steel weld. The results obtained are as follows: (1) Metal suitable for HAZ was made by applying thermal treatment to simulate the thermal hysteresis during welding to clearly define creep and fatigue properties in a HAZ softened zone. (2) Creep strength of the weld metal and the welded joint was equal to that of the base metal. However, the welded joint ruptured in the HAZ softened zone and its ductility decreased after about ...

Recent developments in the CONTAIN-LMR code

YBCO films were fabricated on PLD-CeO{sub 2}/IBAD-Gd{sub 2}Zr{sub 2}O{sub 7}/Hastelloy substrates using the advanced TFA-MOD process. The effective thickness of the CeO{sub 2} buffer layer for obtaining high I{sub c} was investigated in short samples of YBCO films. The CeO{sub 2} buffer layer was epitaxially grown on an IBAD-Gd{sub 2}Zr{sub 2}O{sub 7} template tape with 18 deg. of {delta}{phi} by a reel-to-reel PLD system. The in-plane grain alignment of PLD-CeO{sub 2} buffer layers rapidly improved with the thickness and saturated at a critical thickness of 0.8 {mu}m. The size of CeO{sub 2} grains was about 1 {mu}m at the saturated thickness of {delta}{phi}. YBCO films with the thickness of 1 {mu}m were deposited by the TFA-MOD on the CeO{sub 2} buffer layer with different thickness films. Improvement of the CeO{sub 2} in-plane grain alignment resulted in increase of I{sub c}. The I{sub c} values of 250-290 A were obtained with the CeO{sub 2} ...

2007-10-01

Influence of O{sub 2} and N{sub 2}H{sub 4} on the ECP in high temperature water

Through an international collaborative effort, a special version of the CONTAIN code is being developed for integrated mechanistic analysis of the conditions in liquid metal reactor (LMR) containments during severe accidents. The capabilities of the most recent code version, CONTAIN LMR/1B-Mod.1, are discussed. These include new models for the treatment of two condensables, sodium condensation on aerosols, chemical reactions, hygroscopic aerosols, and concrete outgassing. This code version also incorporates all of the previously released LMR model enhancements. The results of an integral demonstration calculation of a sever core-melt accident scenario are given to illustrate the features of this code version. 11 refs., 7 figs., 1 tab.

1990-08-12

Containment closure time following loss of cooling under shutdown conditions of YGN units 3 and 4

The ECP of construction materials in the water steam circuits of power plants is influenced by many parameters, including: reactions of oxidants, such as O{sub 2} or dissolved copper species; and reactions of reducing species, namely N{sub 2}H{sub 4}. Electrochemical measurements were performed to clarify the role of hydrazine for the open circuit potential in water/steam circuits. Current density electrode potential curves of the electrochemical oxidation of hydrazine and the reduction of oxygen in aqueous solutions were measured as a function of temperature in the range from room temperature to approximately 260{degrees}C. The electrode materials used were platinum, gold and Alloy 800 mod.. In addition, corrosion potentials were measured in water containing oxygen or hydrazine.

1992-12-31

Wolsong 2,3 and 4 fuel channel analysis during a large break loss of coolant accident with loss of ECCS injection

The YGN Units 3 and 4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling. The thermal hydraulic analyses were performed for the five cases of RCS configurations under the worst event scenario, unavailable secondary cooling and no RCS inventory makeup, using the RELAP5/MOD3.2 code to investigate the plant behavior. From the analyses results, times to boil, times to core uncovery and times to core heat up were estimated to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. These data provide useful information to the abnormal procedure to cope with the event. 6 refs., 7 figs., 2 tabs. (Author)

1998-12-31

Integrative facility management via the intranet requires reliable communication. Network and security solutions for multi-site facility management; Integratives Gebaeudemanagement ueber Intranet erfordert gesicherte Kommunikation. Network und Security Konzept fuer das technische Multi-Site-Facility-Management

Wolsong 2,3 and 4 fuel channel analysis during a large break loss of coolant accident with loss of ECCS injection (LOCA/LOECC) is performed to obtain the heat load to moderator. Because the single channel analysis requires the establishment of the safety codes and their input decks, the present study follows the same safety analysis methodology found in FSAR of Wolsong 2,3 and 4. From this work we obtain the safety tools such as CATHENA MOD3.5b/Rev.1 and CHAN-II/A MOD2 codes, and their code modeling in a form of code input deck. The analysis consists of two parts: front-end (blowdown period) and back-end. For the front-end analysis the fuel and pressure tube (PT) temperatures, and PT circumferential strains at the end of front-end as well as fuel channel depressurization are calculated using CATHENA code and used as initial and boundary conditions for back-end analysis. The back-end period under the conditions of prolonged low steam flow is ...

2002-10-01

Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

Since enterprises operating offices and facilities at geographically distributed sites have been establishing an intranet for inter-office communication at the management level, the idea of integrating another intranet for multi-site technological facility management comes naturally. The Aachen-based technology enterprise Eurem for this purpose developed a Network and Security concept with solutions that can be customized for linking a variety of computerized control and monitoring systems. The package includes a modular substation named modUS, which is open to the field side as well as to the management side, is capable of efficiently and reliably performing the facility management functions by way of remote controlling via intranets. (orig./CB) [Deutsch] Nachdem in den Unternehmen mit verteilten Liegenschaften die unterschiedlichsten Buerokommunikationssysteme ueber eigene Netzwerke verbunden werden, liegt es nahe, auch verteilte Gebaeudeautomationssysteme in ...

1999-02-08

Time contour expression of limited range phenomena on stack chart; Jugo chart jo deno kyokuchi gensho jikan contour

The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 ...

Structure of the triplet of low-lying states in sup 101 Mo

Time contour expression of limited range phenomena on stack chart is examined for further improvement on the result of the ultimate interpretation in the seismic reflection survey. The policy is made clear from the beginning that local phenomena are to be discussed, and data prior CMP stacking is interpreted in detail. For this purpose, it is effective to make use of the time contour expression in the midpoint-offset plane simultaneously with the CMP and COP panels. For the review of data prior to CMP stacking, it is convenient to use the CMP (CDP) stacking chart in which the data is arranged methodically. In this chart, all the channels which are crude data prior to stacking are plotted on midpoint-offset coordinates, which plane is called the MOD (Midpoint Offset Domain) panel. Various panels can be chosen unrestrictedly, and their mutual relations can be easily grasped. When data points are given a time axis, they can be expressed in a time contour. Studies are ...

1997-05-27

Structure of the triplet of low-lying states in "1"0"1Mo

The properties of the triplet of low-lying states in {sup 101}Mo have been studied through spectroscopy of the {gamma} radation following thermal neutron capture in {sup 100}Mo and {beta}-decay of {sup 101}Nb and through a measurement of the proton angular distributions in the {sup 100}Mo(d, p) reaction with 14 MeV deuteron energy. The half-lives of the 13.5 keV state and the 57.0 keV 5/2{sup +} state have been measured as 226(7) and 133(7) ns, respectively. These values and the quadrupole/dipole mixing ratios of the 13.5 keV and 43.5 keV transitions yield spin and parity 3/2{sup +} for the 13.5 keV level. The E2 components in the 13.5 (3/2{sup +}->1/2{sup +}) and 43.5 keV (5/2{sup +}->3/2{sup +}) transitions are {le} 8x10{sup -4} and 54(9)%, respectively. The possibility of an additional state near to the 57.0 keV level is discussed. IBFM/PTQM calculations, taking into consideration the transitional character of the {sup 100}Mo boson core, account for the ...

1991-06-01

SCDAP/RELAP5/MOD 3.1 code manual: Interface theory. Volume 1

The properties of the triplet of low-lying states in "1"0"1Mo have been studied through spectroscopy of the #gamma# radation following thermal neutron capture in "1"0"0Mo and #beta#-decay of "1"0"1Nb and through a measurement of the proton angular distributions in the "1"0"0Mo(d, p) reaction with 14 MeV deuteron energy. The half-lives of the 13.5 keV state and the 57.0 keV 5/2"+ state have been measured as 226(7) and 133(7) ns, respectively. These values and the quadrupole/dipole mixing ratios of the 13.5 keV and 43.5 keV transitions yield spin and parity 3/2"+ for the 13.5 keV level. The E2 components in the 13.5 (3/2"+#->#1/2"+) and 43.5 keV (5/2"+#->#3/2"+) transitions are #<=# 8x10"-"4 and 54(9)%, respectively. The possibility of an additional state near to the 57.0 keV level is discussed. IBFM/PTQM calculations, taking into consideration the transitional character of the "1"0"0Mo boson core, account for the electromagnetic-transition and ...

On the natural convection cooling in HANARO (Hi-flux Advanced Neutron Application Reactor). Experiment and RELAP5/KMRR simulation

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of off-site power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume describes the organization and manner of the interface between severe accident models which are resident in the SCDAP portion of the code and hydrodynamic models which are resident in the RELAP5 portion of the code. A description of the ...

1995-06-01

1 2 3 4

Municipal waste combustion study: costs of flue-gas-cleaning technologies. Final report

The natural circulation experiments were conducted to confirm the cooling capability and the flow characteristics of the natural convection in the HANARO (Hi-flux Advanced Neutron Application Reactor). The tests were done at the power levels of 2%, 3% and 4% (1.2MW_t_h) of full power. The flow rates and temperatures at various locations of the primary and secondary cooling loops were measured at each power level. The temperature distributions in the chimney and the pool were also obtained. Through tests, the flow paths of the natural circulation and the cooling capability of the reactor were confirmed as designed. In addition, the simulation for the natural circulation tests was made by using RELAP5/KMRR, which was modified from RELAP5/MOD2 for applying to the HANARO conditions. The simulation results show that RELAP5/KMRR gives reasonable predictions for the flow rate and the coolant temperature during natural circulation condition in the HANARO. (author)

MARS CODE MANUAL VOLUME V: Models and Correlations

This report is an assessment of emission-control costs for municipal-waste combustors (MWCs). The details of the cost estimates, including their development, components, and cost premises, are addressed. A model-plant approach was used in the sizing and costing of the emission control systems. Due to differences in the feed-waste characteristics, combustion parameters, and emissions, separate cost estimates were required for mass burning (MB), modular (MOD), refuse-derived fuel (RDF), and fluid-bed combustion (FBC) type furnaces. Cost estimates were developed for control of particulate matter (PM) emissions only and for control of both acid gas and PM emissions from the MWC model plants. Controlled PM emission levels of 0.03, 0.02, and 0.01 gr/dscf, corrected to 12% CO/sub 2/, and 90 and 70% reductions of HC1 and SO2, respectively, were used to develop the control cost estimates. Costs were developed using the cost information received from a number of ...

1987-06-01

Four loss-of-flow accidents in the SEAFP first wall/blanket cooling system

Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This models and correlations manual provides a complete list of detailed information of the thermal-hydraulic models used in MARS, so that this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the ...

2002-09-01

Determination of parameters of the environment for equipment qualification at the Dukovany NPP. Post-accident parameters on the +14.7 m floor. Operating parameters on the +14.7 m floor and in the hermetic zone. Rev. 4

This report presents the thermal-hydraulic analysis of four Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the alternative SEAFP reactor design. The LOFAs considered result from a loss of electrical power for the recirculation pump in the primary cooling circuit. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analyses, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall and blanket. For the LOFA without plasma shutdown, significant loss of heat removal due to dryout occurs at the midplane of the outboard first wall cooling pipes about 41 s after pump trip. For the three LOFA cases with emergency plasma shutdown that have been studied, the temperature increase in the Be-coating at the midplane of the outboard first wall is limited to about 30 K. (orig.).

1994-07-01

Break Nodalization Influence to IAEA-SPE-4 Test Simulation

A detailed outline of the application of the MELCOR and RELAP5/MOD3.1 codes to the analysis of the thermohydraulic response and determination of other parameters of the medium on the floor is given for several classes of secondary coolant circuit accidents along with the description of the related facilities. An overview is presented of the maximum values and time behavior of the thermohydraulic parameters, pressure, temperature, relative humidity, and water level on the floor. Transverse rupture of the steam generator, main steam header, or main feedwater header piping during normal operation is considered as the initiating event. Pressure is only 10% higher than the atmospheric pressure. Air temperature attains a value as high as 100 degC. Relative humidity is 100%, persisting as long as the steam source is available. The water level is typically about 8 cm and never exceeds 15 cm. (M.D.). 16 tabs., 37 figs., 32 refs.

Advanced PWR technology development -Development of advanced PWR system analysis technology-

A small break LOCA event simulation with no high pressure injection system available, known as International Atomic Energy Agency Standard Problem Exercise no. 4 (IAEA-SPE-4), was performed on the PMK-2 integral test facility in Budapest in 1993. This paper analyses the response of the PMK-2 facility, a model of VVER-440 nuclear power plant, using the latest released version MOD3.2.1.2 of the RELAP5 thermal-hydraulic code. After several years of the SPE-4 experiment analyses, many problems have emerged and been studied. Main goal of the present analyses was to study the main influencing parameters for adequate modelling of the hexagonal core channel with 19-rod bundle and phenomena during the core uncovery. Some influencing parameters have been identified, mostly on the primary side, but some also on the secondary side. This is exact simulation of main coolant pump coast down, hydro-accumulators water temperature and connections to the primary system. Some concern ...

1998-06-15

A procedure for temperature-stress fields calculation of WWER-1000 primary circuit in PTS event

The primary scope of this study is to establish the analysis technology for the advanced reactor designed on the basis of the passive and inherent safety concepts. This study is extended to the application of these technology to the safety analysis of the passive reactor. The study was performed for the small and medium sized reactor and the large sized reactor by focusing on the development of the analysis technology for the passive components. Among the identified concepts the once-through steam generator, the natural circulation of the integral reactor, heat pipe for containment cooling, and hydraulic valve were selected as the high priority items to be developed and the related studies are being performed for these items. For the large sized passive reactor, the study plans to extend the applicability of the best estimate computer code RELAP5/MOD3 which is widely used for the safety analyses of the reactor system. The improvement and supplementation study of ...

1995-07-01

Study of the effect of noncondensable gas on heat transfer phenomena in horizontal steam generator of pactel facility with CATHARE2 V1.5a

The paper presents the procedure of an investigation of WWER-1000 primary circuit temperature-stress field by the use of thermohydraulic computation data for a pressurized thermal shock event ''Core overcooling''. The procedure is based on a model of the plane stress state with ideal contact between wall and medium for the calculation. The computation data are calculated on the base of WWER-1000 thermohydraulic model by the RELAP5/MOD3 codes. This model was developed jointly by the Bulgarian and BNL/USA staff to provide an analytical tool for performing safety analysis. As a result of calculations by codes the computation data for temperature field law (linear laws of a few distinguished parts) and pressure of coolant at points on inner surface of WWER-1000 primary circuit equipment are received. Such calculations can be used as a base for determination of all-important load-carrying sections of the primary circuit pipes and vessels, which need further ...

1997-05-05

Seismic stratigraphy and stratigraphic modelling of the South-eastern German Molasse Basin

Lappeenranta University of Technology (LTKK) and VTT Energy carried out a series of preliminary tests in 1999 to study the behavior of noncondensable (NC) gases in VVER geometry. The tests aimed at studying the effect of NC gases on system thermal-hydraulics and on heat transfer in a horizontal steam generator (HSG). The system behavior can be affected by hydrogen produced in the core in case of a severe accident, by nitrogen from hydro-accumulators released into the primary circuit in case of a loss-of-coolant accident (LOCA) and more generally by any NC gas in all cases where cooling is ensured by natural circulation. A secondary objective of the tests - the first series of tests ever performed with NC gas with PACTEL - was to find out, if the instrumentation of PACTEL was adequate for this type of tests and if it was functioning properly. This paper presents the measured and calculated (CATHARE code version V15a mod 2.1) results of the test NCg-l. It was carried ...

2001-03-20

A study of passive and inherent safety design concepts for advanced light= water reactors

Although the German Molasse Basin can be regarded as a mature hydrocarbon province, no regional sequence stratigraphic analysis has been carried out so far. We have studied seismic lines and well data from the region between the Isar and Inn rivers (SE Germany) that have been generously supplied by German oil companies (DEE, BEB, Mobil, RWE-DEA and Wintershall). Initial work indicates that five major seismic sequences within three main depositional cycles are developed. The Alpine thrust belt to the south serves as the primary sediment source in the foreland basin. However, sedimentary infill mainly took place parallel to the basin axis. Our analysis suggests that the stratigraphic development of the Molasse Basin was mainly controlled by eustatic sea-level changes which caused the shoreline to shift in the W-E direction. The shifting of the depocenter axis in a N-S direction was controlled by the tectonic evolution of the thrust belt. The sea-level curve determined by seismic ...

1995-08-01

A phenomenological model of the thermal hydraulics of convective boiling during the quenching of hot rod bundles

The five thermal-hydraulic concepts chosen for conceptual study of advanced PWR systems have been studied as follows: (1) Critical Heat Flux in passive PWR Conditions: review of previous works (various of correlations, analysis of parametric trends) on CHF, assessment and improvement of CHF prediction models for round tubes, development of the prediction model on bundle CHF with considering the correction factor calculated from the tube data base, design and construction of the intermediate-pressure CHF experimental loop, extension of CHF data base by performing the experiments at low-flow, and low-quality conditions (2) Passive Cooling Concepts for Concrete Containment Systems: Selection of the external condenser by comparing and reviewing between passive cooling concepts for concrete containment system concepts, survey and review of previous studies (theoretical mechanism of condensation heat transfer and effect of non-condensable gases) on the condensation phenomena, design and ...

1997-07-01

A fundamental approach to better understand the lithium insertion mechanisms in electrode materials; Une approche fondamentale pour mieux comprendre les mecanismes d`insertion du lithium dans les materiaux d`electrodes

In this paper, a phenomenological model of the thermal hydraulics of convective boiling in the post-critical-heat-flux (post-CHF) regime is developed and discussed. The model was implemented in the TRAC-PF1/MOD2 computer code (an advanced best-estimate computer program written for the analysis of pressurized water reactor systems). The model was built around the determination of flow regimes downstream of the quench front. The regimes were determined from the flow-regime map suggested by Ishii and his coworkers. Heat transfer in the transition boiling region was formulated as a position-dependent model. The propagation of the CHF point was strongly dependent on the length of the transition boiling region. Wall-to-fluid film boiling heat transfer was considered to consist of two components: first, a wall-to-vapor convective heat-transfer portion and, second, a wall-to-liquid heat transfer representing near-wall effects. Each contribution was considered separately in ...

1983-10-14

A Scheme of 3-D Breakdown-whip Analysis Methodology for High Energy Piping

The development of rechargeable lithium batteries with a high mass capacity, made with non-toxic and low cost materials is an important industrial challenge. Morphological and structural modifications occurring in the electrode materials during charge-output cycles should not lower the electrochemical characteristics and the cycling properties of the battery. Thus the structure of electrode materials must be sufficiently deformable and stable to support the constraints linked with lithium intercalation and de-intercalation (ions and electrons absorption/extraction). The aim of this work is to explain some characteristics (mass capacity, ions and electrons mobility, cycling) using the relation between some mechanisms of lithium insertion (sites occupation, lattice reduction mods) and the nature of atoms and chemical bonds (covalence, ionicity). This approach is developed on 2-D models of crystallized and vitreous sulfur compounds (CdI{sub 2} type) with a large ...

1996-12-31