Use of RELAP50MOD3.3 Code to Get Fluid Dynamic Stability Maps [PDF]
, 2012 The analysis that has been carried out in the present paper shows the possibility to predict the onset of density wave oscillations by means of RELAP5/MOD3.3 code for simple geometry systems, whereas the difference between 1D and 3D approximations is not
De Salve, Mario +2 more
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Improved technique of a complex analysis of crack resistance of WWER-1000 nuclear reactor cold leg nozzle under termal shock. Report 1. Thermo-hydraulic and transient thermal calculations [PDF]
, 2014 В Україні для тепло-гідравлічних розрахунків різних сценаріїв, можливих на реакторних установках АЕС,використається розрахунковий комплекс RELAP5. Проблема полягає в тім, що розрахунки в RELAP5 звичайно проводять у районі патрубка корпуса реактора на ...
Iakovlev, A. +3 more
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RELAP5/MOD3.3 Best Estimate Analyses for Human Reliability Analysis
Science and Technology of Nuclear Installations, 2010 To estimate the success criteria time windows of operator actions the conservative approach was used in the conventional probabilistic safety assessment (PSA). The current PSA standard recommends the use of best-estimate codes.
Andrej Prošek, Borut Mavko
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Computationally efficient stratified flow wet angle correlation for high resolution simulations [PDF]
, 2019 n high resolution two-phase pipe flow simulations, such as slug capturing simulation for liquid-gas pipe flow, explicit calculation of stratified flow wet angle has been proposed to improve computational speed of simulations.
Kara, Fuat, Oloruntoba, O
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EnergiesThe RELAP5 code is a computational tool designed for transient simulations of light water reactor coolant systems under hypothesized accident conditions.
Young Hwan Lee +2 more
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Science and Technology of Nuclear Installations, 2019 COMSOL Multiphysics has been used to conduct thermal-hydraulic analysis in multiple nuclear applications. The aim of this study is to benchmark the prediction accuracy of COMSOL Multiphysics in performing thermal-hydraulic analysis of TRIGA (Training ...
Ahmed K. Alkaabi, Jeffrey C. King
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Simulación del accidente de SGTR en un PWR-W con TRACE para distintas metodologías de análisis determinista de seguridad [PDF]
, 2012 A Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR) can lead to an atmospheric release bypassing the containment via the secondary system and exiting though the Pressurized Operating Relief Valves of the affected Steam Generator ...
Jiménez Varas, Gonzalo +3 more
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Analisi dello shock termico in un RPV tipo WWER100 in condizioni di DEGB [PDF]
, 2005 Nel settore dell'Ingegneria nucleare, è di importanza fondamentale l'analisi di problemi relativi alla sicurezza degli impianti già esistenti od in fase di progettazione.
Frustaci, Luca
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二次开发的
He jishu为验证经二次开发后的RELAP5程序在倾斜条件下的适用性，基于缩小比例的反应堆二次侧非能动余热排出系统（Passive Residual Heat Removal System，PRS）实验装置，开展了倾斜角度为-24°~+24°的工况下的余热排出实验。在此基础上，使用二次开发后的RELAP5程序对实验工况进行了模拟计算，利用实验数据验证其适用性，并进行了进一步对比分析。研究结果表明：二次开发后的RELAP5程序能有效预测倾斜条件下系统运行特性的变化 ...
郝 承明 +8 more
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Computation of Gap Conductance in Different Fuel Assemblies in VVER-1000 Type Reactors [PDF]
مجله علوم و فنون هستهای, 2011 In this paper, a calculation for fresh fuels gap conductance at different axial lengths of fuel assemblies of the VVER-1000 type reactors has been made using two models of Calza-Bini and Relap5.
M Rahgoshay, KH Shokri
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