Results 71 to 80 of about 5,558 (186)
Transient analysis of OSU-MASLWR with RELAP5
Journal of Physics: Conference Series, 2022 Abstract The present paper deals with the assessment of the original and a modified version of RELAP5/MOD3.3 against the OSU Multi Application Small Light Water Reactor (OSU-MASLWR). The new implemented features regard suitable correlations for the heat transfer coefficient evaluation in helical geometry.
Molinari M. +3 more
openaire +1 more source
Thermo-hydraulic analysis of the windowless target system [PDF]
, 2008 The target system, whose function is to supply an external neutron source to a subcritical core in order to sustain the neutron chain reaction, is the most critical part of an ADS being subject to severe thermo-mechanical loading and material damage due
Bianchi, Fosco +2 more
core +1 more source
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
Science and Technology of Nuclear Installations, 2017 This paper presents a faster solver named NRLU (Node Reordering Lower Upper) factorization solver to improve the solution speed for the pressure equations, which are formed by RELAP5/MOD3.3. The NRLU solver uses the oriented graph method and minimal fill-
Li Ge, Wei Liu, Jianqiang Shan
doaj +1 more source
N Reactor RELAP5 model benchmark comparisons [PDF]
, 1988 This report documents work performed at the Idaho National Engineering Laboratory (INEL) in support of Westinghouse Hanford Company safety analyses for the N Reactor. The portion of the work reported here includes comparisons of RELAP5/MOD2-calculated data with measured plant data for: (1) a plant trip reactor transient from full power operation; and ...
Fletcher, C. D., Bolander, M. A.
openaire +2 more sources
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
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
Opposing Mixed Convection Heat Transfer for Turbulent Single‐Phase Flows
International Journal of Energy Research, Volume 2024, Issue 1, 2024.Convection, wherein forced and natural convections are prominent, is known as mixed convection. Specifically, when a forced convection flow is downward, this flow is called opposing flow. The objectives of this study are to gain a comprehensive understanding of opposing flow mixed convection heat transfer and to establish the prediction methodology by ...
Kosuke Motegi +5 more
wiley +1 more source
Development and Assessment of the COBRA/RELAP5 Code. [PDF]
Journal of Nuclear Science and Technology, 1997 The COBRA/RELAP5 code has been developed to combine the realistic three-dimensional reactor vessel model of COBRA-TF with RELAP5/MOD3.2, thus to produce an advanced system analysis code with a multidimensional thermal-hydraulic module. This paper provides the integration scheme of the two codes. The results of developmental assesments are also provided,
Jae-Jun JEONG, Suk Ku SIM, Sang Yong LEE
openaire +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