A Design study of high electric power for fast reactor cooled by super critical light water

Koshizuka, Seiichi*

JNC-TJ9400 2000-011, 102 Pages, 2000/03

JNC-TJ9400-2000-011.pdf:2.71MB

In order to evaluate the possibility to achieve high electric power by a fast reactor with supercritical light water, the design study was carried out on a large fast reactor core with high coolant outlet temperature (SCFR-H). Since the reactor coolant circuit uses once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure, it is possible to design much simpler and more compact reactor systems and to achieve higher thermal efficiency than those of current light water reactors. The once-through direct cycle system is employed in current fossil-fired power plants. In the present study, three types of core were designed. The first is SCFR-H with blankets cooled by ascending flow, the second is SCFR-H with blankets cooled by descending flow and the third is SCFR-H with high thermal power. Every core was designed to achieve the thermal efficiency over 43%, positive coolant density reactivity coefficient and electric power over 1600MW. Core characteristics of SCFR-Hs were compared with those of SCLWR-H (electric power: 1212MW), which is a thermal neutron spectrum reactor cooled and moderated by supercritical light water, with the same diameter of the reactor pressure vessel. It was shown that SCFR-H could increase the electric power about l.7 times maximally. From the standpoint of the increase of a reactor thermal power, a fast reactor has advantages as compared with a thermal neutron reactor, because it can increase the power density by adopting tight fuel lattices and eliminating the moderator region. Thus, it was concluded that a reactor cooled by supercritical light water could further improve the cost competitiveness by using a fast neutron spectrum and achieving a higher thermal power.

Analysis of natural circulation charactristic in middle size ATR

*; *

PNC-TJ9381 93-001, 158 Pages, 1993/02

PNC-TJ9381-93-001.pdf:2.12MB

Present nuclear power reactor adopts a forced circulation system. But, the natural circulation studies for light water reactor are recently preceeding. So, on purpose to examine possibility of ATR natural circulation reactor in 1000Mwt class, we evaluated the natural flow and core cooling characteristic for ATR. In this study, the sensitivity analyses for the natural circulation characteristic by change inlet pipe diameter, outlet pipe diameter, core length, and downcomer pipe height were executed. Then the sensitivities are as follows. (1)Influence of inlet pipe diameter. From analyses of the case which was executed using 3B as inlet pipe diameter, against using 2B in the base case, it was clarified that channel flow of the case using 2B is more stable than that of the another. (2)Influence of outlet pipe diameter. From analyses of the cases which were executed using 4B and 5B as outlet pipe diameter, against using 3B in the base cases, it was clarified that the cases using pipes of a large diameter have more natural flow than the other cases, because of decreasing pressure loss at outlet pipes. (3)Influence of core length. From analyses of the cases which were executed using 3.2m and 2.7m as core length, against using 3.7m in the base case, it was clarified that the cases using short core length have more natural flow than the other cases, because of decreasing pressure loss at the core. (4)Influence of downcomer height. From analyses of the cases which were executed using 20m and 30m as downcomer against using 15m in the base case, it was clarified that the cases using high downcomer height have more natural flow than the other cases, because of increasing driving head which is the diferential pressure between the core and the downcomer. And, from the analyses results of (1) to (4), it was crarified that the item of the most largest sensitivity was downcomer height. And then, using the combination which have most natural circulation flows for ...

The Effect of damage on fatigue crack growth rate under high-temperature creep-fatigue multiplication

*; Kaji, Yoshiyuki; *; *

Nippon Kikai Gakkai Rombunshu, A, 57(542), p.2349 - 2354, 1991/10

https://doi.org/10.1299/kikaia.57.2349

no abstracts in English