Induced nuclear reactions: processes – systems – and elements – Reactor structures – Circulating fluid within reactor
Reexamination Certificate
2001-07-13
2003-02-25
Carone, Michael J. (Department: 3641)
Induced nuclear reactions: processes, systems, and elements
Reactor structures
Circulating fluid within reactor
C376S221000
Reexamination Certificate
active
06526115
ABSTRACT:
BACKGROUND OF THE INVENTION
This invention is related to supercritical water cooled nuclear reactors and electric power generation plants utilizing such nuclear reactors.
In the prior art, pressurized water nuclear reactors (PWRs) and boiling water nuclear reactors (BWRs) are well known and commercially operated. A typical PWR power plant comprises steam generators outside of the reactor, which are a kind of heat exchangers where heat is transferred from the primary coolant which has been heated in the PWR. The secondary coolant is changed into steam in the steam generators. The steam is used to rotate steam turbines and then generate electricity. Typical pressure in a PWR reactor vessel is about 15 MPa and the primary coolant temperature at the outlet of the PWR is about 320° C. Typical pressure of the secondary coolant at the outlets of the steam generators is about 7 MPa, and the efficiency of the typical PWR electric power generation is about 35 percent.
On the other hand, a typical BWR power plant does not have an external steam generator. Steam for steam turbines is generated in the BWR itself, and the steam at the outlet of the BWR has a pressure of about 7 MPa and a temperature of about 290° C. Thus, the efficiency of the typical BWR electric power generation is about 35 percent.
Supercritical-pressure water nuclear reactors have been proposed as shown in
FIG. 1
which is similar to those disclosed in Japanese Patent Application Publication (Tokkai-Hei) 8-313664, the disclosure of which is hereby incorporated by reference in its entirety.
Referring to
FIG. 1
, a nuclear reactor vessel
22
contains a core
23
where nuclear reaction occurs. The core
23
heats up supercritical-pressure water into a supercritical-fluid having a pressure of about 25 MPa and a temperature of about 450° C. Water having a pressure and a temperature above the critical point values that are 22.1 MPa and 374° C., respectively, is not liquid nor steam, but behaves like steam. Therefore, that kind of fluid may be called “steam” or “supercritical-pressure steam” hereinbelow.
The supercritical-pressure steam is sent to a steam turbine
7
via a main steam line
24
. The steam turbine
7
is rotated by the steam and drives an electric power generator
8
. The steam is then condensed into water in a condenser
9
. The condensed water is then pumped up to supercritical pressure by a feed water pump
26
and sent back to the nuclear reactor vessel
22
via a feed water line
25
.
The efficiency of the supercritical water cooled reactor electric power generation is about 40 percent which is higher than that of PWRs and BWRs due to the improved steam condition supplied to the steam turbine. However, the steam condition of the supercritical water cooled reactor is still lower than that of supercritical-pressure thermal power plants.
Furthermore, the reactor vessel of the supercritical water cooled reactor should have thicker walls due to the higher pressure compared to PWRs and BWRs. The wall thickness for a supercritical water cooled reactor may be 1.7 times of that for a PWR with a same diameter of the reactor vessel.
SUMMARY OF THE INVENTION
Accordingly, it is an object of the present invention to provide an improved supercritical water cooled reactor and an improved electric power generation plant utilizing such a nuclear reactor, which are more efficient and economical.
There has been provided, in accordance with an aspect of the present invention, a supercritical water cooled reactor comprising: a reactor vessel including: a shell part for containing sub-critical pressure coolant, and an end part for containing supercritical-pressure coolant which is separated from the sub-critical pressure coolant in the reactor vessel; a core-support plate with a plurality of through-holes, the core-support plate disposed in and fixed to the reactor vessel so that the core-support plate divides space inside the reactor vessel into a supercritical-pressure portion and a sub-critical pressure portion; a plurality of fuel tubes with both open ends fixed to the through-holes, the open ends being communicated to the supercritical-pressure portion, outside of the fuel tubes being disposed in the sub-critical pressure portion; a plurality of nuclear fuel assemblies disposed in the fuel tubes; means for introducing supercritical-pressure water into the supercritical-pressure portion; means for extracting supercritical-pressure steam generated in the fuel tubes out of the supercritical-pressure portion; means for introducing sub-critical pressure coolant into the sub-critical pressure portion; means for extracting sub-critical pressure coolant out of the sub-critical pressure portion; a plurality of control rods which are arranged so that the control rods can be inserted into the sub-critical pressure portion adjacent to the fuel tubes through the shell part; and a control rod drive for driving the control rods from outside of the reactor vessel.
There has also been provided, in accordance with another aspect of the present invention, an electric power generation plant having: (a) a supercritical water cooled reactor comprising: a reactor vessel including: a shell part for containing sub-critical pressure coolant, and an end part for containing supercritical-pressure coolant which is separated from the sub-critical pressure coolant in the reactor vessel; a core-support plate with a plurality of through-holes, the core-support plate disposed in and fixed to the reactor vessel so that the core-support plate divides space inside the reactor vessel into a supercritical-pressure portion and a sub-critical pressure portion; a plurality of fuel tubes with both open ends fixed to the through-holes, the open ends being communicated to the supercritical-pressure portion, outside of the fuel tubes being disposed in the sub-critical pressure portion; a plurality of nuclear fuel assemblies disposed in the fuel tubes; means for introducing supercritical-pressure water into the supercritical-pressure portion; means for extracting supercritical-pressure steam generated in the fuel tubes out of the supercritical-pressure portion; means for introducing sub-critical pressure coolant into the sub-critical pressure portion; means for extracting sub-critical pressure coolant out of the sub-critical pressure portion; a plurality of control rods which are arranged so that the control rods can be inserted into the sub-critical pressure portion adjacent to the fuel tubes through the shell part; and a control rod drive for driving the control rods from outside of the reactor vessel; (b) a higher pressure turbine receiving the supercritical-pressure steam extracted from the supercritical-pressure portion of the reactor; (c) means for extracting part of output steam of the higher pressure turbine to introduce the output steam to the sub-critical pressure portion of the reactor; (d) a lower pressure turbine receiving the sub-critical pressure coolant extracted from the sub-critical pressure portion of the reactor; and (e) an electric generator driven by at least on of the higher and lower pressure turbines.
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patent: 05-333184 (1993-12-01), None
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Kazuyoshi Kataoka et al., “Neutronic Feasibility of Supercritical Steam Cooled Fast Breeder Reactor”, Journal of Nuclear Science and Technology, vol. 28, No. 6, Jun. 1991, pp. 585-587.
Fueki Eiko
Kataoka Kazuyoshi
Ookawa Masahiro
Carone Michael J.
Kabushiki Kaisha Toshiba
Matz Daniel
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