3 edition of Assessment of PWR steam generator modelling in RELAP5/MOD2 found in the catalog.
Assessment of PWR steam generator modelling in RELAP5/MOD2
by Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Available from Supt. of Docs., U.S. G.P.O., National Technical Information Service [distributor in Washington, DC, Springfield, VA
Written in English
|Statement||prepared by J.M. Putney, R.J. Preece.|
|Series||International agreement report -- NUREG/IA-0106.|
|Contributions||Preece, R. J., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research., National Power, Technology, and Environmental Centre (Great Britain)|
|The Physical Object|
|Pagination||ix, 111 p.|
|Number of Pages||111|
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Abstract. 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 Cited by: 2.
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.
Assessment of PWR steam genera Staff View; Cite this; Text this; Email this; Export Record. Export to EndNoteWeb; Export to EndNote; Save to List; Add to Book Bag Remove from Book Bag. Saved in: Assessment of PWR steam generator modelling in RELAP5/MOD2 / a Assessment of PWR steam generator modelling in RELAP5/MOD2 / |c prepared by J.
Assessment of RELAP5/MOD against a main steam isolation valve closure at TRILLO I Nuclear Power Plant / by: Lucas, A. de, et al. Published: () Assessment of RELAP5/MOD3/V5m5 against the UPTF test no.
11 (countercurrent flow in PWR hot leg) / by: Curca-Tivig, F., Published: (). Buy Assessment of PWR steam generator modelling in RELAP5/MOD2 (SuDoc Y 3.N /) on FREE SHIPPING on qualified orders. Assessment of PWR steam generator modelling in RELAP5/MOD2 (OCoLC) Material Type: Government publication, National government publication, Internet resource: Document Type: Book, Internet Resource: All Authors / Contributors: J M Putney; R J Preece; U.S.
Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Technical Report: Assessment of PWR Steam Generator modelling in RELAP5/MOD2.
International Agreement Report. Buy Assessment of PWR steam generator modelling in RELAP5/MOD2 (SuDoc Y 3.N /) by J. Putney (ISBN:) from Amazon's Book Store. Everyday low prices and free delivery on eligible : J.
Putney. Modeling Two-Phase Flow in the Downcomer of a Once-Through Steam Generator using RELAP5/MOD2. Randy Raymond Clark, Jr. Thesis submitted to the faculty of Virginia Polytechnic Institute and State University in partial fulfillment of the requirements for the degree of.
Master of Science. Mechanical Engineering. Alan A. Kornhauser, Chair. RELAPD is a simulation tool that allows users to model the coupled behavior of the reactor coolant system and the core for various operational transients and postulated accidents that might occur in a nuclear (Reactor Excursion and Leak Analysis Program) can be used for reactor safety analysis, reactor design, simulator training of operators, and as an educational tool by License: Proprietary.
Download Citation | CFD Simulations in Complex Nuclear Containment Configurations | Two-phase flows with water droplets have a significant influence on the thermal-hydraulic behaviour within. One of the major effects is deterioration of heat transfer at steam generator, mainly in the condensation regime.
After deep analysis of condensation models in RELAP5 computer code, a set of. RELAP5/MOD2 RELAP5/MOD2 is a system thermal hydraulics code which solves simultaneously the separate conservation equations of mass momentum and energy for steam and water in a twophase flow. Also solved is a field equation enabling calulation of boron concentration in.
A few remarkable results taken from the SB-LOCA counterpart test database (e.g., see D'Auria et al.,Blinkov et al.,D'Auria and Galassi, ) are reported in Fig. Fig. Fig. for the system pressure, water mass inventory, and rod surface temperature.
These figures were produced by University of Pisa. The objective is to give an impression of the counterpart test Cited by: 4. Marko Čepin, Modification of the main feedwater fault trees due to steam generator replacement and power uprate: Contribution to integrated safety assessment of NPP Krško modernisation project.
Rev. 2, (IJS delovno poročilo, ), Ljubljana, Institut "Jožef Stefan", US PWR steam generator management: An overview. SciTech Connect. Welty, C.S. This paper provides an overview on the status of steam generator management activities.
Modeling and Simulation of U-tube Steam Generator. NASA Astrophysics Data System (ADS) Zhang, Mingming; Fu, Zhongguang; Li, Jinyao; Wang, Mingfei. The U-tube natural c. This is paper for thermal-hydraulic.
Helmholtz-Zentrum Dresden Rossendorf. Toggle navigation. Twitter; LinkedIn; YouTube; Helmholtz-Netwerk. A steam explosion has a precise definition in the parlance of light water reactor safety. It is generally defined as a physical interaction between a hot and a cold liquid, in the case ofLWR materials, chemical interactions may also occur.
Available Models Figure Convective Boiling Factor, F Figure Nucleate Boiling Suppression Factor, S Figure Parametric Effect of CHF with Variation of Initial and Boundary Conditions Figure Burnout in Cross Flow over Tubes Figure Critical Quality Compared to Proportional Power to Outer Surface Figure Comparison of "Local" versus "Integral" Hypotheses for CHF Figure /5(2).The most extensive effort to date has taken place in the United States.
In large part, this concern about water hammer can be traced back to the early 's when the number of reported water hammer occurrences in U.S.
nuclear power plants was increasing dramatically .5/5(1).Nuclear Power - System Simulations and Operation Fig. Importance analysis of plant status parameters Instead of applying a full scoped BELOCA methodology to cover both model and plant status uncertainties, a deterministic- realistic hybrid methodology (DRHM) was developed to support LOCA licensing analysis.