Cadmium Transport through Molten Salts in the Reprocessing of Spent Fuel for the Integral Fast Reactor

1993 ◽  
Vol 102 (3) ◽  
pp. 331-340 ◽  
Author(s):  
K. Michael Goff ◽  
Alfred Schneider ◽  
James E. Battles
Keyword(s):  
Fast Reactor ◽  
Molten Salts ◽  
Spent Fuel ◽  
Cadmium Transport

scholarly journals Study on CANDLE Burning Fast Reactor with Initial Core Using Plutonium from LWR Spent Fuel

2017 ◽  
Vol 131 ◽  
pp. 15-20 ◽  
Author(s):  
Hiroki Osato ◽  
Jun Nishiyama ◽  
Toru Obara
Keyword(s):  
Fast Reactor ◽  
Spent Fuel

scholarly journals Thermal-hydraulics of internally heated molten salts and application to the Molten Salt Fast Reactor

2014 ◽  
Vol 501 ◽  
pp. 012030 ◽  
Author(s):  
Carlo Fiorina ◽  
Antonio Cammi ◽  
Lelio Luzzi ◽  
Konstantin Mikityuk ◽  
Hisashi Ninokata ◽  
...  
Keyword(s):  
Molten Salt ◽  
Fast Reactor ◽  
Molten Salts ◽  
Thermal Hydraulics

scholarly journals Decay Heat of Fast Reactor Spent Fuel.

1999 ◽  
Vol 41 (9) ◽  
pp. 946-953 ◽  
Author(s):  
Takafumi AOYAMA ◽  
Tadahiko TORIMARU ◽  
Akihiro YOSHIDA ◽  
Yoshio ARII ◽  
Soju SUZUKI
Keyword(s):  
Fast Reactor ◽  
Spent Fuel ◽  
Decay Heat

scholarly journals Minor actinides transmutation in equilibrium cores of next generation FRs

2019 ◽  
Vol 5 (4) ◽  
pp. 353-359
Author(s):  
Alexander V. Egorov ◽  
Yurii S. Khomyakov ◽  
Valerii I. Rachkov ◽  
Elena A. Rodina ◽  
Igor R. Suslov
Keyword(s):  
Isotopic Composition ◽  
Fast Reactor ◽  
Fuel Cycle ◽  
Synergetic Effect ◽  
Transient State ◽  
Advanced Technology ◽  
Thermal Reactor ◽  
Spent Fuel ◽  
Operating Period ◽  
Nitride Fuel

The Russian Federation is developing a number of technologies within the «Proryv» project for closing the nuclear fuel cycle utilizing mixed (U-Pu-MA) nitride fuel. Key objectives of the project include improving fast reactor nuclear safety by minimizing reactivity changes during fuel operating period and improving radiological and environmental fuel cycle safety through Pu multi-recycling and МА transmutation. This advanced technology is expected to allow operating the reactor in an equilibrium cycle with a breeding ratio equaling approximately 1 with stable reactivity and fuel isotopic composition. Nevertheless, to reach this state the reactor must still operate in an initial transient state for a lengthy period (over 10 years) of time, which requires implementing special measures concerning reactivity control. The results obtained from calculations show the possibility of achieving a synergetic effect from combining two objectives. Using МА reprocessed from thermal reactor spent fuel in initial fuel loads in FR ensures a minimal reactivity margin during the entire fast reactor fuel operating period, comparable to the levels achieved in equilibrium state with any kind of relevant Pu isotopic composition. This should be combined with using reactivity compensators in the first fuel micro-campaigns. In the paper presented are the results of simulation of the overall life cycle of a 1200 MWe fast reactor, reaching equilibrium fuel composition, and respective changes in spent fuel nuclide and isotopic composition. It is shown that МА from thermal and fast reactors spent fuel can be completely utilized in the new generation FRs without using special actinide burners.

Possible routes for pyrochemical separation: Focus on the reductive extraction in fluoride media

2012 ◽  
Vol 85 (1) ◽  
pp. 71-87 ◽  
Author(s):  
Sylvie Delpech
Keyword(s):  
High Temperature ◽  
Extraction Method ◽  
Molten Salts ◽  
Experimental Measurements ◽  
Spent Fuel ◽  
Essential Point ◽  
Pure Compounds ◽  
Spent Fuel Reprocessing ◽  
Chloride Salts ◽  
Fuel Reprocessing

Molten salts (MSs) such as fluoride or chloride salts at high temperature (400–800 °C) are solvents known for their high solvation power and electroactivity range. Rare earths, lanthanides, actinides, and refractory metals can be dissolved, treated, and purified in MSs. The properties of these solvents are particularly interesting for nuclear spent-fuel reprocessing. The pyrochemical separation and extraction of solutes can be performed using several methods taking into account the effects of redox and/or acidity. This paper is focused on the reductive extraction method performed by contacting a liquid metal (LM) containing reductive species and an MS. The analytical model developed to calculate the efficiency of such a method is detailed in this paper. To apply this model, one essential point is the establishment of a database related to the redox and solvation properties of solutes in MSs. The approach retained to propose a database based on the analysis of both thermochemical data of pure compounds and experimental measurements reported in the literature is described in this paper in the case of lanthanides in fluoride MSs. The use of the database to calculate efficiency as a function of process parameters is given in this paper as well as the comparison between two reducing agents considered.

scholarly journals Preliminary proliferation study of the molten salt fast reactor

2020 ◽  
Vol 6 ◽  
pp. 5 ◽  
Author(s):  
Michel Allibert ◽  
Elsa Merle ◽  
Sylvie Delpech ◽  
Delphine Gerardin ◽  
Daniel Heuer ◽  
...  
Keyword(s):  
Molten Salt ◽  
Fast Reactor ◽  
Molten Salts ◽  
Fuel Cycle ◽  
Power Reactor ◽  
Nuclear Material ◽  
Reactor System ◽  
Molten Salt Reactor ◽  
International Forum ◽  
Large Power

The molten salt reactor designs, where fissile and fertile materials are dissolved in molten salts, under consideration in the framework of the Generation IV International Forum, present some unusual characteristics in terms of design, operation, safety and also proliferation resistance issues. This paper has the main objective of presenting some proliferation challenges for the reference version of the Molten Salt Fast Reactor (MSFR), a large power reactor based on the thorium fuel cycle. Preliminary studies of proliferation resistance are presented here, dedicated to the threat of nuclear material diversion in the MSFR, considering both the reactor system itself and the processing units located onsite.

Initial core design of CANDLE burning fast reactor using plutonium from LWR spent fuel

2018 ◽  
Vol 120 ◽  
pp. 501-508 ◽  
Author(s):  
Hiroki Osato ◽  
Jun Nishiyama ◽  
Toru Obara
Keyword(s):  
Fast Reactor ◽  
Spent Fuel
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