Results 71 to 80 of about 5,521 (168)
Analyzing the influence of an increase in the thermal power of energy generating unit at the nuclear power plant on the behavior of beyond the design basis accident [PDF]
, 2018 Учитывая потенциальную возможность для энергоблоков АЭС Украины увеличения тепловой мощности реакторной установки до 104 % от проектного значения, которая реализована в ряде стран, с одной стороны и уроки, извлеченные из аварии на АЭС Фукусима-1 с другой,
Никуленков, Анатолий Геннадьевич +2 more
core +1 more source
Worst‐Case Accident Analysis of Accident Tolerant Fuel in NuScale Using RELAP5/MOD3‐Based Code
International Journal of Energy Research, Volume 2024, Issue 1, 2024.Research in nuclear engineering focusses on improving the safety of light water reactors (LWRs), driven by accidents like Fukushima in 2011. The severity of this accident was a result of active cooling system failures and cladding material oxidation resulting in hydrogen explosions.
Willem Zuidersma +3 more
wiley +1 more source
Core Flow Distribution from Coupled Supercritical Water Reactor Analysis
Science and Technology of Nuclear Installations, 2014 This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady ...
Po Hu, Paul P. H. Wilson
doaj +1 more source
Dynamic PRA: an Overview of New Algorithms to Generate, Analyze and Visualize Data [PDF]
, 2013 State of the art PRA methods, i.e. Dynamic PRA (DPRA) methodologies, largely employ system simulator codes to accurately model system dynamics. Typically, these system simulator codes (e.g., RELAP5 ) are coupled with other codes (e.g., ADAPT, RAVEN
A. Yilmaz +10 more
core
Science and Technology of Nuclear Installations, 2010 The Institute of Neutron Physics and Reactor Technology (INR) is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability.
V. Sánchez +3 more
doaj +1 more source
Nuclear Engineering and Technology, 2018 An experiment was performed for the OECD/NEA ROSA-2 Project with the large-scale test facility (LSTF), which simulated a steam generator tube rupture (SGTR) accident due to a double-ended guillotine break of one of steam generator (SG) U-tubes with ...
Takeshi Takeda
doaj +1 more source
Simulation of rod ejection accident byPARCS code [PDF]
, 2015 This paper describes reactor core model used for simulating REA. The model was designed in PARCS utilizing graphical interface SNAP. The data for model were given from benchmark NEACPR L-335.
Matějková, J.
core
Analysis of Density Wave Oscillations in Helically Coiled Tube Once-Through Steam Generator
Science and Technology of Nuclear Installations, 2016 Helically coiled tube Once-Through Steam Generator (H-OTSG) is one of the key equipment types for small modular reactors. The flow instability of the secondary side of the H-OTSG is particularly serious, because the working condition is in the range of ...
Junwei Hao +7 more
doaj +1 more source
Hot Zero and Full Power Validation of PHISICS RELAP-5 Coupling [PDF]
, 2013 PHISICS is a reactor analysis toolkit developed over the last 3 years at the Idaho National Laboratory. It has been coupled with the reactor safety analysis code RELAP5-3D.
A., Epiney +4 more
core
RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation
Science and Technology of Nuclear Installations, 2018 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 ...
Andrej Prošek, Marko Matkovič
doaj +1 more source