Results 1 to 10 of about 2,211 (141)
RELAP5/MOD2 models and correlations [PDF]
, 1988 A review of the RELAP5/MOD2 computer code has been performed to assess the basis for the models and correlations comprising the code. The review has included verification of the original data base, including thermodynamic, thermal-hydraulic, and geothermal conditions; simplifying assumptions in implementation or application; and accuracy of ...
R.A. Dimenna +7 more
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, 2016 The paper presents applicability of built-in RELAP5/MOD3.2 cladding deformation model for VVER-1000 fuel with cladding of Zr+1 % Nb alloy. Experimental data and simplified model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the hot channel blockage after cladding swelling and rupture in the ...
Yu. Vorobyov +2 more
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Numerical Simulations with RELAP5-3D and RELAP5/mod3.3 of the Second Experimental Campaign on In-Box LOCA Transients for HCLL TBS [PDF]
Energies, 2021 In-box LOCA was identified as one of the worst accidental scenarios for the HCLL TBS (Helium Cooled Lithium-Lead Test Blanket System). Aiming to experimentally analyze the consequences of this transient, ENEA designed and built THALLIUM (Test HAmmer in Lead LIthiUM), a facility that reproduces the LiPb loop of the HCLL TBS.
Alessandro Venturini +2 more
exaly +4 more sources
Uncertainty Analysis for RELAP5-3D [PDF]
, 2011 In its current state, RELAP5-3D is a 'best-estimate' code; it is one of our most reliable programs for modeling what occurs within reactor systems in transients from given initial conditions. This code, however, remains an estimator. A statistical analysis has been performed that begins to lay the foundation for a full uncertainty analysis.
USDOE Office of Nuclear Energy (NE) +2 more
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SCDAP/RELAP5/MOD 3.1 code manual: Interface theory. Volume 1 [PDF]
, 1995 The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light-water-reactor coolant systems during severe accidents. The newest version of the code is SCDAP/RELAP5/MOD3. The US Nuclear Regulatory Commission (NRC) decided that there was a need for a broad technical review of the code by recognized experts to determine overall ...
E.W. Coryell, C.M. Allison, G.A. Berna
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Study on the Density Wave Instability in Natural Circulation System under Rolling Conditions
International Journal of Energy Research, Volume 2023, Issue 1, 2023., 2023 In the ocean environment, the thermal fluid system installed on the ship moves with the waves. This unsteady transient motion imposes additional body forces on the liquid. Thereby, the thermohydraulic properties of the thermal fluid system are changed, especially for natural circulation systems with a low driving force.
Houjun Gong +6 more
wiley +1 more source
Science and Technology of Nuclear Installations, Volume 2023, Issue 1, 2023., 2023 When the feedwater valve at the outage loop of the floating nuclear power plant leaks, thermal stratification occurs in the steam generator. It causes lower water temperature in the outage loop. The extent of hazard of this phenomenon cannot be directly determined by the existing measurement parameters, which poses a threat to the operational safety of
Xinxin Liu +4 more
wiley +1 more source
Science and Technology of Nuclear Installations, Volume 2023, Issue 1, 2023., 2023 This paper summarises the results of an experimental campaign carried out at SIET on the ELSMOR facility built in 2022 to validate a decay heat removal system for the E‐SMR. Based on the passive mechanisms of natural circulation, the system aims to dissipate the reactor decay heat to a water pool, using two heat exchangers: a plate‐type one coupling ...
Roberta Ferri +8 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,
JEONG, Jae-Jun +2 more
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Simulations of Core Damage Progression for TMI‐2 Severe Accident Using CINEMA Computer Code
Science and Technology of Nuclear Installations, Volume 2023, Issue 1, 2023., 2023 As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. The CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant ...
Rae-Joon Park +6 more
wiley +1 more source