Decay heat removal analyses on the heavy liquid metal cooled fast breeding reactor; Comparisons of the decay heat removal characteristics on Lead, Lead-Bismuth and Sodium cooled reactors

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.

A Study on advanced marine reactor

*

Nihon Zosen Gakkai-Shi, 0(734), p.511 - 521, 1990/08

http://ci.nii.ac.jp/naid/110003865358

no abstracts in English

MHD Eguilibrium Analysis of JAERI Experimental Fusion Reactor

*

JAERI-M 8413, 37 Pages, 1979/09

JAERI-M-8413.pdf:0.79MB

no abstracts in English

Preliminary Safety Analysis of the Reactor Cooling System for Tokamak Experimental Fusion Reactor

JAERI-M 7772, 109 Pages, 1978/08

JAERI-M-7772.pdf:2.88MB

no abstracts in English

First Preliminary Design of an Experimental Fusion Reactor

JAERI-M 7300, 545 Pages, 1977/09

JAERI-M-7300.pdf:15.12MB

no abstracts in English