Zobrazeno 1 - 10
of 10
pro vyhledávání: '"Thomas K. S. Liang"'
Publikováno v:
Nuclear Technology. 169:50-60
The limiting blowdown event for the design of an advanced boiling water reactor (ABWR) containment shifts from a conventional recirculation line break to a feedwater line break (FWLB) by implementing reactor internal pumps. As a result, coupled blowd
Publikováno v:
Nuclear Technology. 166:146-155
In the innovative design of the advanced boiling water reactor (ABWR), conventional recirculation loops are removed and replaced by multiple reactor internal pumps. Therefore, there is no major penetration of the reactor pressure vessel (RPV) below t
Publikováno v:
Nuclear Technology. 153:184-196
The blowdown of feedwater (FW) line breaks (FWLBs) has been successfully analyzed by using the Appendix K version of RELAP5-3D. To adequately simulate a feedwater blowdown event, one must consider the main steam system, the turbine system, the moistu
Publikováno v:
Nuclear Technology. 139:233-252
In light water reactors, particularly the pressurized water reactor (PWR), the severity of a loss-of-coolant accident (LOCA) would limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the gre
Publikováno v:
Nuclear Technology. 136:292-300
With the consideration of mass unbalance, coolant shrinking, and compressibility, a model for reactor coolant leakage evaluation has been developed to quantify on-line the system leakage rate with conventional system measurements, regardless of where
Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis
Publikováno v:
Nuclear Technology. 136:37-49
The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluatio
Autor:
Thomas K. S. Liang, Richard R. Schultz
Publikováno v:
Nuclear Technology. 133:355-358
In light water reactors, particularly the pressurized water reactors, the severity of loss-of-coolant accidents (LOCAs) will limit how high the reactor power can extend. Although the best-estimate LOCA methodology can provide the greatest margin on t
Autor:
Fu-Kuang Ko, Thomas K. S. Liang
Publikováno v:
Nuclear Technology. 129:13-25
Although only a few percent of residual power remains during plant outages, the associated risk of core uncovery and corresponding fuel overheating has been identified to be relatively high, partic...
Autor:
Thomas K. S. Liang
Publikováno v:
Nuclear Power-System Simulations and Operation
The Loss of Coolant Accident (LOCA) is one of the most important design basis accidents (DBA). In light water reactors, particularly the pressurized water reactor (PWR), the severity of a LOCA will limit how high the reactor power can operate. In the
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::37abc718fd783fdf894e60cbcca6d310
http://www.intechopen.com/articles/show/title/development-of-an-appendix-k-version-of-relap5-3d-and-associated-deterministic-realistic-hybrid-meth
http://www.intechopen.com/articles/show/title/development-of-an-appendix-k-version-of-relap5-3d-and-associated-deterministic-realistic-hybrid-meth
Publikováno v:
10th International Conference on Nuclear Engineering, Volume 4.
As most of the nuclear power plants, on-site spent fuel pools (SFP) of Taiwan’s plants were not originally designed with a storage capacity for all the spent fuel generated over the operating life by their reactors. For interim spent fuel storage,