1,721,213 research outputs found
Towards a more realistic MELCOR model for a dry cask for spent nuclear fuel. Part I: sensitivity studies
No full textThe United States (US) Department Of Energy (DOE) has addressed the thermal analysis of the Spent Nuclear Fuel (SNF) stored within a dry cask system as a matter of high priority. In this regard, it is of utmost importance that simulation tools effectively reproduce the general thermal behavior of the modelled cask, including heat exchange and removal. Temperature distribution in the different components of the system is usually the focus of performed thermal analyses. In particular, attention is paid to the maximum temperature reached in the fuel cladding, namely the Peak Cladding Temperature (PCT). Within this framework, the present paper is the first of a two-paper series aimed at developing a more accurate model for the HI-STORM 100S cask. The dry cask in question is modelled and its behavior is simulated by means of the MELCOR code (version 2.2.18019). Stressing the need for a more realistic model rather than a conservative one, this paper reports the efforts undertaken to evaluate the influence of some specific modelling choices on the PCT. The study of the cask performance is therefore conducted taking into consideration three main factors: the axial power distribution in the Fuel Assembly (FA), the flow losses in the air gap between the internal canister and the external overpack, and the conductivity of the overpack concrete
Towards a more realistic MELCOR model for a dry cask for spent nuclear fuel. Part II: application
No full textNowadays, a great deal of attention is devoted to the development of best-estimate models able to produce more realistic outcomes. This is also the case for system codes, such as MELCOR, that are being mostly used in a conservative way especially when dealing with the licensing process. The above-mentioned need for more realistic results is at the core of this two-paper series related to the creation of a more accurate MELCOR model for the HI-STORM 100S dry cask. The findings obtained from the sensitivity studies carried out in the Part I are leveraged to set up an improved MELCOR model, the characteristics of which are consistent with the typical features of Spent Nuclear Fuel (SNF), and with geometrical and material properties of the cask itself. The addition of an axial power profile in the Fuel Assembly (FA), the better characterization of the flow losses in the air gap between internal metallic canister and external concrete-based overpack, and the choice of an appropriate value for the concrete thermal conductivity, are taken into account conjointly in this Part II. The outcomes from the improved MELCOR simulation are reported mainly in terms of the Peak Cladding Temperature (PCT), being the variable under regulatory surveillance. However, in addition to PCT, calculated temperature profiles are displayed and compared against the ones resulting from the previous model
OECD-CSNI Benchmark problem 2 on combustible gas distribution in a large dry containment case DA: calculation results using the FUMO computer code
No full textOsservazioni sulla biologia di Solea vulgaris Quens. della classe d'eta' O+ in un'area costiera della Sardegna
No full textOECD-CSNI ISP N.23: Rupture of a large diameter pipe Within the HDR containment. Blind post-test calculation using the FUMO computer code
No full textERMSAR 2019 conference of NUGENIA TA2/SARNET on research on severe accidents in nuclear power plants
Full text linkThe 9th ERMSAR conference, which selected papers are gathered in this Annals of Nuclear Energy special issue, was hosted by the ÚJV Řež in Prague (Czech Republic) from 18 to 20 March 2019. The Scientific Programme Committee involved eight researchers from diverse organizations (CEA, CIEMAT, ENEA, IAEA, IRSN, JSI and University of Pisa). The conference gathered 163 participants from 23 countries and 73 organizations (19% of participants came from out of Europe such as Canada, Egypt, India, Japan, USA, China, and the Republic of Korea), confirming its status as a major international event in nuclear reactor safety. Sixty-five papers plus several posters were presented in the 3 days and significant time was allocated after each presentation for questions and open discussions
OECD/CSNI ISP-35 on NUPEC Hydrogen Mixing and Distribution Test-7-1: UNIPI Final Calculations with the FUMO Code
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