Results 51 to 60 of about 1,536,451 (138)
Bayesian Uncertainty Quantification of Reflooding Model With PSO–Kriging and PCA Approach
Science and Technology of Nuclear Installations, Volume 2025, Issue 1, 2025.To improve the process of best estimate plus uncertainty (BEPU) for nuclear safety assessment and calibration of thermal–hydraulic models for error reduction, inverse uncertainty quantification (IUQ) is proposed in recent years to quantify the uncertainty of model parameters in reactor program.
Ziyue Zhang +4 more
wiley +1 more source
SCDAP/RELAP5/MOD3 code development [PDF]
, 1992 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.
Siefken, J. L. +2 more
core
Accident Analysis in Research Reactors [PDF]
The first Research Reactor reaches its criticality on 2nd December 1942. Nowadays, more than 300 research reactors are currently in operation. Over the past fifty years, research reactors have progressed through a variety of tasks.
Adorni, Martina
core +1 more source
Science and Technology of Nuclear Installations, 2015 In earlier study we have demonstrated that RELAP5 can predict flow instability parameters (flow rate, oscillation period, temperature, and pressure) in single channel tests in CIRCUS-IV facility. The main goals of this work are to (i) validate RELAP5 and
Viet-Anh Phung, Pavel Kudinov
doaj +1 more source
Experimental and Numerical Results of LIFUS5/Mod3 Series E Test on In-Box LOCA Transient for WCLL-BB
Energies, 2021 The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design of the WCLL-BB (water-cooled lead-lithium breeding blanket).
Marica Eboli +5 more
doaj +1 more source
International Journal of Energy Research, Volume 2024, Issue 1, 2024.In the case of multiple‐failure accidents in light water reactors, it is imperative to demonstrate that a significant fuel damage can be prevented via prearranged mitigation measures performed by operators. However, the time elapsed from the initiation of the event to the implementation of mitigation actions significantly influences the progress and ...
Jia Yu +3 more
wiley +1 more source
ICONE-22450 RELAP5-3D CODE APPLICATION FOR RBMK-1500 REACTOR CORE ANALYSIS [PDF]
, 2020 The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important.
Evaldas Bubelis +2 more
core
Yuanzineng kexue jishuAfter Fukushima nuclear accident, the passive concept is widely adopted in the thermal safety design of advanced nuclear reactors. The in-pool passive residual heat removal system (PRHRS) can also provide security guarantee for the domestic reactor ...
LIAN Qiang1, 2, 3, ZHU Longxiang1, 2, 3, TANG Simiao1, 2, 3, HUANG Tao4, ZHANG Yong4, PAN Liangming1, 2
doaj +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
MNSR transient analysis using the RELAP5/Mod3.2 code
Nuclear Engineering and Technology, 2020 To support the safe operation of the Miniature Neutron Source Reactor (MNSR), a thermo-hydraulic transient model using the RELAP5/Mod3.2 code was simulated.
S. Dawahra, K. Khattab, F. Alhabit
doaj +1 more source