Zobrazeno 1 - 10
of 40
pro vyhledávání: '"Francesco D'Auria"'
Analysis of loss of cooling accident in VVER-1000/V446 spent fuel pool using RELAP5 and MELCOR codes
Publikováno v:
Nuclear Engineering and Technology, Vol 55, Iss 8, Pp 3102-3113 (2023)
Following the Fukushima nuclear disaster, the simulation of accidents in the spent fuel pool has become more noticeable. Despite the low amount of decay heat power, the consequences of the accidents in a spent fuel pool (SFP) can be severe due to the
Externí odkaz:
https://doaj.org/article/8e59b7bad11b42029aabe4988dc6f5e8
Autor:
Francesco D’Auria, Romney B. Duffey
Publikováno v:
Nuclear Energy and Technology, Vol 8, Iss 2, Pp 77-90 (2022)
After three quarters of a century using nuclear fission to produce energy, Nuclear Reactor Safety and Risk constitutes an established technological sector. A key feature is continuous updating following new discoveries and progress in knowledge, resu
Externí odkaz:
https://doaj.org/article/43a6d4041264496a8737aba831b1df9f
Publikováno v:
Nuclear Energy and Technology, Vol 5, Iss 3, Pp 183-199 (2019)
The present paper deals with the proposal of an additional safety barrier for the class of large (1000 MWe or more) Light Water Reactors (LWR) now in operation, in construction, or under design. Emphasis is given to the motivations or the needs for t
Externí odkaz:
https://doaj.org/article/25ec36fb5372470987574b4ed4ab354d
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2013 (2013)
The present paper deals with the analytical study of the PKL experiment G3.1 performed using the TRACE code (version 5.0 patch1). The test G3.1 simulates a fast cooldown transient, namely, a main steam line break. This leads to a strong asymmetry cau
Externí odkaz:
https://doaj.org/article/64224f28145041268c3185345172486d
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2012 (2012)
The paper deals with the evaluation of the Fukushima-Daiichi Nuclear Power Plant (NPP) accident in Units 1 to 4: an attempt is made to discuss the scenario within a technological framework, considering precursory documented regulations and predictabl
Externí odkaz:
https://doaj.org/article/24d81c6524d04e82985d6a5afa421826
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2011 (2011)
Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel
Externí odkaz:
https://doaj.org/article/02a30a5490ea44b1bed3c7b81f7fdb7f
Autor:
Francesco D'Auria, Oscar Mazzantini
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2011 (2011)
Within the licensing process of the KWU Atucha II PHWR (Pressurized Heavy Water Reactor), the BEPU (Best Estimate Plus Uncertainty) approach has been selected for issuing of the Chapter 15 on FSAR (Final Safety Analysis Report). The key steps of the
Externí odkaz:
https://doaj.org/article/ad8ea9e33e554b4eb214670db96adc2f
Autor:
Dino Araneo, Paolo Ferrara, Fabio Moretti, Andrea Rossi, Andrea Latini, Francesco D'Auria, Oscar A. Mazzantini
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2011 (2011)
The present paper describes the main features and an application to a real Nuclear Power Plant (NPP) of an Integrated Software Environment (in the following referred to as “platform”) developed at University of Pisa (UNIPI) to perform Pressurized
Externí odkaz:
https://doaj.org/article/5be6ef0ff05f40acb84daee9285ccadd
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2009 (2009)
The paper deals with the presentation of the Reliability Evaluation of Passive Safety System (REPAS) methodology developed by University of Pisa. The general objective of the REPAS is to characterize in an analytical way the performance of a passive
Externí odkaz:
https://doaj.org/article/73c1b7288fd2472fbfed1c2cf649f391
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2008 (2008)
Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. The computer code user represents a source
Externí odkaz:
https://doaj.org/article/e55f2b51c2674ae4a112cb40c6199cb9