Zobrazeno 1 - 10
of 46
pro vyhledávání: '"Pham Nhu Viet Ha"'
Publikováno v:
Nuclear Engineering and Technology, Vol 53, Iss 7, Pp 2095-2103 (2021)
A small power ADS design using thorium oxide and diluent matrix reprocessed fuel is proposed for a high transmutation rate, small reactivity swing, and strong safety features. Two fuel matrices (CERCER and CERMET) and different recycled fuel composit
Externí odkaz:
https://doaj.org/article/840393b3d87b479abc4e1ef6bb299ce5
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2019 (2019)
Thermal reactors have been considered as interim solution for transmutation of minor actinides recycled from spent nuclear fuel. Various studies have been performed in recent decades to realize this possibility. This paper presents the neutronic feas
Externí odkaz:
https://doaj.org/article/1efbbfe8d7124b84b4f6fe9f24a07deb
Autor:
Giang Phan, Hoai-Nam Tran, Kien-Cuong Nguyen, Viet-Phu Tran, Van-Khanh Hoang, Pham Nhu Viet Ha, Hoang Anh Tuan Kiet
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2017 (2017)
Neutronics analysis has been performed for the 500 kW Dalat Nuclear Research Reactor loaded with highly enriched uranium fuel using the SRAC code system. The effective multiplication factors, keff, were analyzed for the core at criticality conditions
Externí odkaz:
https://doaj.org/article/1a986050c1b44025a1b29fa2c620d69e
Publikováno v:
In Annals of Nuclear Energy April 2022 168
Publikováno v:
In Annals of Nuclear Energy April 2019 126:110-121
Publikováno v:
Journal of Nuclear Engineering and Radiation Science; 20240101, Issue: Preprints p1-38, 38p
Akademický článek
Tento výsledek nelze pro nepřihlášené uživatele zobrazit.
K zobrazení výsledku je třeba se přihlásit.
K zobrazení výsledku je třeba se přihlásit.
Publikováno v:
VNU Journal of Science: Mathematics - Physics. 37
This paper presents the steady-state analysis results of the OECD/NEA and U.S. NRC PWR MOX/UO2 (MOX: Mixed Oxide) Core Transient Benchmark with the modern MCNP6 Monte Carlo code based on the ENDF/B-VII.1 evaluated nuclear data library. The purpose wa
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::5cf8dcf41cb1edd38779609eafe97346
https://doi.org/10.25073/2588-1124/vnumap.4621
https://doi.org/10.25073/2588-1124/vnumap.4621
Publikováno v:
VNU Journal of Science: Mathematics - Physics. 37
In this study, the SCALE/TRITON code (based on deterministic method) and the Serpent 2 code (based on Monte Carlo method) were utilized to prepare the group constants of the pressurized water reactor (PWR) mixed-oxide (MOX) fuel assemblies for transi
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::7f71aa1157d27fa1a6e9cfdfbe289879
https://doi.org/10.25073/2588-1124/vnumap.4596
https://doi.org/10.25073/2588-1124/vnumap.4596
Publikováno v:
VNU Journal of Science: Mathematics - Physics. 37
The ADS (accelerator driven system) is recognized as a promising system to annihilate the radioactivity of nuclear waste with its inherent safety feature and waste transmutation potential. Thus, conceptual designs of ADS are widely carrying out. In o
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::ae5a59a1774fc59edbf3f4ea062a4b82
https://doi.org/10.25073/2588-1124/vnumap.4506
https://doi.org/10.25073/2588-1124/vnumap.4506