scholarly journals Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR). Metallic Fuels Program

1993 ◽  
Author(s):  
G.L. Batte ◽  
R.G. Pahl ◽  
G.L. Hofman
Keyword(s):  
Fast Reactor ◽  
Current Status ◽  
Metallic Fuels

Current Status of the Development of the Fast Reactor Fuel Cycle in India

2009 ◽  
pp. 120-126
Author(s):  
K.V. Govindan Kutty ◽  
P.R. Vasudeva Rao ◽  
Baldev Raj
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Reactor Fuel ◽  
Current Status ◽  
Fast Reactor Fuel

Current Status of Fast Reactor Critical Experiments

1971 ◽  
Vol 13 (10) ◽  
pp. 582-597
Author(s):  
Shoji NOMOTO ◽  
Tsutomu IIJIMA
Keyword(s):  
Fast Reactor ◽  
Current Status

scholarly journals Fuel performance evaluation of annular metallic fuels for an advanced fast reactor concept

2019 ◽  
Vol 352 ◽  
pp. 110157 ◽  
Author(s):  
Yinbin Miao ◽  
Nicolas Stauff ◽  
Aaron Oaks ◽  
Abdellatif M. Yacout ◽  
Taek K. Kim
Keyword(s):  
Performance Evaluation ◽  
Fast Reactor ◽  
Fuel Performance ◽  
Metallic Fuels

scholarly journals An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

2017 ◽  
Vol 9 (12) ◽  
pp. 2225 ◽  
Author(s):  
Wuseong You ◽  
Ser Hong
Keyword(s):  
Burning Rate ◽  
Fast Reactor ◽  
Reactor Core ◽  
Core Concept ◽  
Metallic Fuels

scholarly journals Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

2013 ◽  
Vol 2013 ◽  
pp. 1-18 ◽  
Author(s):  
Yeong-il Kim ◽  
Yong Bum Lee ◽  
Chan Bock Lee ◽  
Jinwook Chang ◽  
Chiwoong Choi
Keyword(s):  
Fast Reactor ◽  
Electricity Generation ◽  
Nuclear Power ◽  
Large Scale ◽  
Development Program ◽  
Energy Resources ◽  
Test Facility ◽  
Current Status ◽  
Hydraulic Test ◽  
Design Concepts

Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR) design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

Static and transient analysis of a medium-sized sodium cooled fast reactor loaded with oxide, nitride, carbide and metallic fuels

2016 ◽  
Vol 87 ◽  
pp. 761-771 ◽  
Author(s):  
Youpeng Zhang ◽  
Konstantin Mikityuk
Keyword(s):  
Fast Reactor ◽  
Transient Analysis ◽  
Metallic Fuels

Development of Structural Codes for JSFR Based on the System Based Code Concept

2014 ◽  
Author(s):  
Tai Asayama ◽  
Takashi Wakai ◽  
Masanori Ando ◽  
Satoshi Okajima ◽  
Yuji Nagae ◽  
...  
Keyword(s):  
Research And Development ◽  
Fast Reactor ◽  
Structural Reliability ◽  
Reliability Evaluation ◽  
Current Status ◽  
Japan Society ◽  
Ongoing Research ◽  
Reactor Types ◽  
Conceptual Study ◽  
Leak Before Break

This paper overviews the current status of the ongoing research and development as well as activities for codification of structural codes for the Japan Sodium Cooled Fast Reactor (JSFR), the demonstration fast reactor which is in the phase of conceptual study. Not only the design and construction code which has been published and updated on a regular basis, codes on welding, fitness-for-service, leak-before-break evaluation as well as the guidelines for structural reliability evaluation are being developed. The basic strategy for the development is to fully take advantage of the favorable technical characteristics associated with sodium-cooled fast reactors; the codes will be developed based on the System Based Code concept, a concept that materializes code rules that are most suitable to the reactor types they are applied to. The above mentioned set of codes are planned to be published from the Japan Society of Mechanical Engineers in 2016.

Current Status and Development of Core Calculation Methodology for Sodium Cooled Fast Reactor in Mitsubishi

2010 ◽  
Author(s):  
Kazumi Ikeda ◽  
Hiroyuki Moriwaki ◽  
Wataru Nakazato
Keyword(s):  
Fast Reactor ◽  
Physical Phenomenon ◽  
Hydraulic Calculation ◽  
Current Status ◽  
Monte Carlo Code ◽  
Design Codes ◽  
Calculation Methodology ◽  
Calculation Results ◽  
Dimensional Cell ◽  
Calculation Code

This paper presents current status and development of nuclear calculation methodology for sodium cooled fast reactor in Mitsubishi and two core design codes; an original two dimensional cell/lattice calculation code, PIJHEX and a thermal hydraulic calculation code, MIX-MKII. Mitsubishi Atomic Power Industries, Inc. developed several core design codes for Monju and JOYO and recently this activity continues still further for the next FBR in Japan. It is explained in this paper that the PIJHEX is confirmed to be valid from the comparison of Monte Carlo code and the examination on physical phenomenon in calculation results. Besides, the approach toward design and development of code system in Mitsubishi and MIX-MKII are introduced briefly.

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