Research and Development Issues for Fast Reactor Structural Design Standard (FDS)

2004 ◽  
Author(s):  
Naoto Kasahara ◽  
Masanori Ando ◽  
Masaki Morishita ◽  
Hiroshi Shibamoto ◽  
Yoshihiko Tanaka ◽  
...  
Keyword(s):  
Research And Development ◽  
Fast Reactor ◽  
Structural Design ◽  
Structural Integrity ◽  
Failure Criteria ◽  
Current Status ◽  
Design Standard ◽  
Design Concepts ◽  
Component Design ◽  
Inelastic Design

For the realization of safe and economical fast reactor (FR) plants, the Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) are cooperating on a research project titled “Feasibility Study on Commercialized FR Cycle Systems.” To certify the design concepts and validate their structural integrity, the research and development of the “Fast Reactor Structural Design Standard (FDS)” is recognized as being an essential theme. FDS considers the general characteristics of FRs and the design requirements for their rationalization. Three main problem areas related to research and development issues were identified by FDS. The first is “refinement of failure criteria,” which takes characteristic design conditions into account. The next is the development of “guidelines for inelastic design analysis” in order to predict the elastic plastic and creep behaviors of high-temperature components. Finally, efforts are being made toward preparing “guidelines for thermal load modeling” for FR component design where thermal loads are dominant. Their research plans and current status are explained.

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.

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.

Main R&D issues for fast reactor structural design standard

2008 ◽  
Vol 238 (2) ◽  
pp. 287-298 ◽  
Author(s):  
Naoto Kasahara ◽  
Kyotada Nakamura ◽  
Masaki Morishita ◽  
Hiroshi Shibamoto ◽  
Kazuhiko Inoue
Keyword(s):  
Fast Reactor ◽  
Structural Design ◽  
Design Standard

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

The Current Status of Research and Development on Selection and Classification of Enlisted Personnel

1990 ◽  
Author(s):  
Jesse Orlansky ◽  
Frances Grafton ◽  
Clessen J. Martin ◽  
William Alley ◽  
Bruce Bloxom
Keyword(s):  
Research And Development ◽  
Current Status ◽  
Enlisted Personnel

Critical Section Selection Methodology for the U.S. EPR™ Standard Nuclear Power Plant

2010 ◽  
Author(s):  
Se-Kwon Jung ◽  
Adam Goodman ◽  
Joe Harrold ◽  
Nawar Alchaar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Design ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Evaluation ◽  
Critical Section ◽  
Selection Methodology ◽  
Critical Sections ◽  
The U.S

This paper presents a three-tier, critical section selection methodology that is used to identify critical sections for the U.S. EPR™ Standard Nuclear Power Plant (NPP). The critical section selection methodology includes three complementary approaches: qualitative, quantitative, and supplementary. These three approaches are applied to Seismic Category I structures in a complementary fashion to identify the most critical portions of the building whose structural integrity needs to be maintained for postulated design basis events and conditions. Once the design of critical sections for a particular Seismic Category I structure is complete, the design for that structure is essentially complete for safety evaluation purposes. Critical sections, taken as a whole, are analytically representative of an “essentially complete” U.S. EPR™ design; their structural design adequacy provides reasonable assurance of overall U.S. EPR™ structural design adequacy.

Sign in / Sign up

Export Citation Format

Share Document