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JNC TN4400 2000-002, 33 Pages, 2000/06
JNC-TN4400-2000-002.pdf:5.22MB
An on-site plant analyzer can provide analysis support in evaluating plant dynamic characteristics when unplanned events occur in a nuclear power station. The plant analyzer contains a plant-dynamics analysis code, which efficiently and quickly analyzes the plant dynamic characteristics. Elements being developed for the on-site plant analyzer include utilities to build plant models for performing analyses and to retrieve plant operation data. The addition of these elements to the analysis code supports the plant-dynamics analysis works in MONJU, in particular, to assist the analyses of start up tests. The system contains the FBR plant-dynamics analysis code "Super-COPD", which is based on the "COPD" code that was used in the safety licensing of MONJU. One feature of the system is that all operations, e.g., assembling plant models for analysis, are prepared using a GUI (Graphical user Interface). In addition to this feature, the system is able to retrieve directly on- and off-line plant data from MIDAS, the Monju Integrated Data Acquisition System. These plant data are used to supply time-dependent boundary conditions for the plant analysis models. For this report, two case studies were performed. First, the analysis result of a turbine trip test at 40% power operation using the full plant model is described. For the second, performance of the IHX model was evaluated using retrieved plant data for boundary conditions. With the development of this system, improvement in the efficiency of analyses of MONJU start-up tests is expected.
Sakai, Takaaki; *; Ohshima, Hiroyuki; Yamaguchi, Akira
JNC TN9400 2000-033, 94 Pages, 2000/04
JNC-TN9400-2000-033.pdf:4.36MB
The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. ln this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube fairer accidents in a steam generator. ln this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in "Equivalent plant" with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. ln conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to conform the heat transfer reduction by the oxidize film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance.