Code development and characteristics analysis for Leak Before Break in the pipelines of Sodium-cooled Fast Reactor

2019 ◽  
Vol 133 ◽  
pp. 777-794 ◽  
Author(s):  
Di Wu ◽  
Minyang Gui ◽  
Jing Zhang ◽  
Yingwei Wu ◽  
Chenglong Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Characteristics Analysis ◽  
Leak Before Break ◽  
Code Development

Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor

Kerntechnik ◽  
2018 ◽  
Vol 83 (3) ◽  
pp. 232-236 ◽  
Author(s):  
D. L. Zhang ◽  
P. Song ◽  
S. Wang ◽  
X. Wang ◽  
J. Chen ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Cooling System ◽  
Pool Type ◽  
Code Development

scholarly journals Structural Materials and Code Development for Japanese Sodium-Cooled Fast Reactors

2014 ◽  
Author(s):  
Tai Asayama ◽  
Yugi Nagae ◽  
Takashi Wakai ◽  
Kazuyuki Tsukimori ◽  
Masaki Morishita
Keyword(s):  
Life Cycle ◽  
Elevated Temperature ◽  
Fast Reactor ◽  
Structural Design ◽  
Reliability Assessment ◽  
Japan Society ◽  
Code System ◽  
Fast Reactors ◽  
Leak Before Break ◽  
Integrated Code

This paper describes the latest status on the development of elevated temperature materials and structural codes for Japanese sodium-cooled fast reactors (SFRs). Based on the extensive research and development activities in the last decades in Japan, two materials, 316FR and Modified 9Cr-1Mo steels were recently incorporated into the 2012 Edition of Fast Reactor Design and Construction Code of the Japan Society of Mechanical Engineers (JSME). Structural design methodologies are continuously being improved towards the next major revision planed in 2016 Edition where methodologies for a 60-year design of Japanese demonstration fast reactor will be provided. Codes and guidelines for fitness-for-service, leak-before-break evaluation and reliability assessment are concurrently being developed utilizing the System Based Code concept aiming at establishing an integrated code system that encompasses a life cycle of SFRs. Paper published with permission.

Development of LBB Assessment Method for Japanese Sodium Cooled Fast Reactor (JSFR) Pipes: 3—Fracture Assessment Methods Considering Compliance at a Crack Part

2010 ◽  
Author(s):  
Takashi Wakai ◽  
Hideo Machida ◽  
Shinji Yoshida ◽  
Yasuhiro Enuma ◽  
Tai Asayama
Keyword(s):  
Thermal Expansion ◽  
Internal Pressure ◽  
Fast Reactor ◽  
Load Balance ◽  
Assessment Method ◽  
Assessment Methods ◽  
Piping System ◽  
Thin Wall ◽  
Fracture Assessment ◽  
Leak Before Break

This paper presents a fracture assessment methods used in leak before break (LBB) assessment of sodium piping system in the Japanese sodium cooled fast reactor (JSFR). Use of thin wall pipes and compact layout of piping system are features of the design in JSFR. Since the internal pressure of piping of JSFR is low, the critical load is thermal expansion. Supposing a through wall crack (TWC) in such piping, the stiffness of the crack part will decrease, the load balance of the piping system will change from the condition without crack. The fracture assessment methods paying attention to this stiffness change at the crack part were proposed and these methods enabled rational LBB assessment. The proposed methods are much effective to loosen LBB conditions for the piping system of which the compliance is low. These methods applied to the LBB assessment of the piping system of JSFR which has the compact layout, and it was checked that the validity of these methods to loosen the LBB conditions.

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.

A code development for leak before break (LBB) leakage from supercritical to subcritical conditions

2018 ◽  
Vol 103 ◽  
pp. 217-228 ◽  
Author(s):  
J. Zhang ◽  
R.H. Chen ◽  
M.J. Wang ◽  
W.X. Tian ◽  
Y.W. Wu ◽  
...  
Keyword(s):  
Leak Before Break ◽  
Code Development

Three-Dimensional Leak before Break Analysis of Nuclear Pipes Containing Cracks

2015 ◽  
Vol 750 ◽  
pp. 376-381
Author(s):  
Wen Fu Liang ◽  
Tong Liu ◽  
Min Shan Liu
Keyword(s):  
High Temperature ◽  
Fast Reactor ◽  
Waste Heat ◽  
Three Dimensional ◽  
Design Guidelines ◽  
Structural Safety ◽  
Piping System ◽  
Pipe Material ◽  
Crack Behavior ◽  
Leak Before Break

Three-dimensional crack behavior simulation analysis and anti-fracture design have been a main subject in fracture theory and engineering application. Piping system is a key part of nuclear power engineering. Utilizing the three-dimensional finite element analysis software ANSYS and the specialized crack analysis programs Franc3D, three-dimensional crack behavior and leak before break (LBB) case were simulated and evaluated of a pipe with a crack in waste heat exhaust system of China Experimental Fast Reactor ( CEFR ). In fast reactor, the piping is working under a high temperature. Therefore, the code RCC-MR.A16 was adopted that is suitable for materials and structural safety design at high temperature. Material used in this article is modified 9Cr1Mo-T91/P91. The analysis model of pipe section was built in three-dimensional entity structure containing a cracks and the high temperature and creep effects were considered. The simulation results show that creep contributes more effect on crack growth than fatigue. The evaluation results on LBB of studied T91 steel pipe with a crack-like defect can satisfy the need of LBB design guidelines. The research results can be referenced in pipe material choose, safety assessment and structural integrity evaluation of a pipe containing defects at high temperature in a fast reactor design.

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