As well known in the art, exemplary nuclear reactors in South Korea include System Plus 80 from Combustion Engineering (CE)(e.g. Hanbit Nuclear Power Plant (NPP) Units 3 and 4), Korean Standard Nuclear Reactors (e.g. Hanul NPP Units 3, 4, 5, and 6; Hanbit NPP Units 5 and 6), OPR1000 (e.g. new Kori NPP Units 1 and 2), APR1400 (new Kori Units 3 and 4, new Ulgin Units 1 and 2, UAE NPP Units 1 to 4), 2-loop Pressurized Water Reactor, etc.
Such NPPs (referred hereinafter simply to as “Nuclear Reactors”) include reactor coolant systems (RCSs) in reactor containment buildings as also disclosed in Korean Patent No. 10-1473665 entitled “Tubing Support Apparatus for Replacement of Parts of Nuclear Power Plant”.
The RCS includes a reactor containing atomic piles and at least one heat-transfer circuit connected thereto.
The circuit includes a steam generator and at least one coolant pump that circulates coolant between the steam generator and the reactor.
In addition, the circuit includes a pressurizer that allows the temperature and pressure of the coolant to be kept constant.
A first large-diameter pipe or a hot leg is connected to one side of the reactor and one side of a suction part of a coolant chamber in the steam generator so as to transmit the coolant, which is heated with contact with a core of the reactor, to the steam generator.
Further, a circulation pipe called a cross-over leg connects one side of an outlet of the coolant chamber and one side of an inlet of a swirl chamber in the coolant chamber.
A cold leg connects the swirl chamber in the coolant chamber and the reactor. The coolant that is cooled at the steam generator and drawn by the coolant pump is transmitted to the reactor via the circulation pipe and the cold leg to cool the core.
In such a nuclear reactor, the steam generator 1 has a typical substructure as shown in FIG. 1.
That is, the steam generator 1 of the nuclear reactor 1 includes a stay cylinder 10 that is maintained at high temperature and a cylindrical skirt 20 that supports the stay cylinder from a lower section thereof.
The skirt 20 is fixedly supported at the bottom thereof by a sliding base 30 by means of a plurality of stud bolts 40. The sliding base 30 is supported by a plurality of (e.g. four) semi-spherical sliders 52, which is provided on a forged bolted plate 50 so as to accommodate a slight motion occurring during the operation of a nuclear reactor.
In this structure, the stud bolt 40 has a conventional solid bolt structure.
During operation, the steam generator produces high temperature heat that, when transmitted to the lower side sliding base 30, elevates the temperature of the sliding base 30 as shown in a heat analysis thermal distribution diagram of FIG. 2a. 
The heat analysis thermal distribution diagram shows that, as a result of analysis with respect of an insulation state, an operation temperature, material, heat transfer, an air flow around a steam generator, etc, a dead air region 70 defined by the skirt 20 and the sliding base 30 is heated to high temperature of up to 131° C. through convection and radiation of high temperature heat (300° C. or more) from the steam generator 1. It could be seen that in such a high temperature state, the skirt 20 was subjected to thermal deformation (e.g. thermal expansion of up to 2.4 mm).
The thermal deformation of the sliding base 30 also causes serious problems as follows.
The thermal deformation, such as thermal expansion, of the sliding base 30 causes restriction to a free motion of the sliding base 30 or interference with an upper surrounding structure of the steam generator 1, resulting in structural vibration of the steam generator 1 and the coolant pump.
The structural vibration causes pipe wearing and vibration stress to the steam generator 1 and after a long operation time, causes material fatigue of small-diameter pipes that have borate embrittlement due to accumulated borate, leading to leakage of a boric acid solution.
Further, as a construction factor of the plumbing in a nuclear reactor, due to weld contraction occurring by a final connection welding between an intermediate pipe and a steam generator nozzle, residual stress remains in the sliding base and a vertical support for a pump. Due to the residual stress, subsidence of the sliding base 30 occurs in response to the weld contraction. As a result, after installation of the steam generator and after final connection welding, the sliding base generally sinks by about 1 mm or less.
Such subsidence of the sliding base may be considered as residual stress of the sliding base 30, and it increases friction force with respect to the sliding base at an initial operating stage of a nuclear reactor, interfering with a transverse sliding motion of the sliding base.
Further, thermal deformation of the sliding base also causes the operating steam generator to be tilted, which may lead to misalignment of parts even after cold shutdown of the steam generator.
Consequently, if the sliding base is deformed so as to be inclined, the steam generator is accordingly tilted so that deformation and interference occur to support structures for the coolant pump and the steam generator, thereby further increasing vibration stress of a nuclear reactor.
It is reported from many countries that such vibration stress causes wear of pipes of the steam generator and of internal components of the coolant pump, and fatigue failure of tubing connected to the RCS during operation of nuclear power plants.
If a nuclear reactor is operated for a long period of time in such condition, a leak may occur from a mechanical seal of a coolant pump and small-diameter pipes in the nuclear reactor.
Accordingly, there is a need to develop a technique to prevent thermal deformation of the sliding base 30 of a nuclear reactor and resultant wear of small pipes of the steam generator and coolant system in a nuclear reactor occurring due to structural vibration of the nuclear reactor.
The applicant has proposed three solutions to address this problem.
A first solution is an air-circulation sleeve 80 that is installed through the center of a sliding base 30 supporting a steam generator to naturally circulate air using a venture effect. This enables a dead air region 70 defined by a stay cylinder and a skirt 20 of the steam generator 1 to be cooled to effectively prevent high temperature heat from being transferred to the sliding base 30 from the steam generator 1, thereby preventing vibrations of a nuclear reactor due to thermal expansion of the sliding base 30.
A second solution is an air-circulating shim plate (not shown) between the bottom of the skirt 20 supporting the steam generator 1 and an engaging surface of the sliding base to allow ambient air to be introduced into a dead air region 70 defined by a stay cylinder and a skirt 20 of the steam generator 1 to be cooled to effectively prevent high temperature heat from being transferred to the sliding base 30 from the steam generator 1, thereby preventing vibrations of a nuclear reactor due to thermal expansion of the sliding base 30.
A third solution is a heat insulation support plate 90 that is closely attached to a lower portion of the steam generator near the stay cylinder and the skirt. The heat insulation support plate is composed of a heat insulation material, a heat shield panel, and a plurality of rigid pieces. The heat insulation support plate is thus prevented from sagging due to operating vibration of a nuclear reactor. The heat insulation support plate serves to effectively prevent high temperature heat from being transferred to the sliding base 30 through the dead air region 70, thereby preventing vibrations of a nuclear reactor due to thermal expansion of the sliding base 30.
The above solutions proposed by the applicant contributes to suppression of thermal deformation of the sliding base 30, having a great effect of stable operation of a nuclear reactor and improvement in lifecycle of equipment of the nuclear reactor.
However, the above solutions are techniques that are applicable at the time of replacement of a steam generator after one cycle of a commercial operation of a nuclear reactor.
Thus, the solutions are difficult to be applied to normally operated nuclear reactor. Further, since even in the operated nuclear reactor, damage of equipment such as a heat pipe or the like can be prevented only when causes of thermal deformation are previously removed, there is a great need to develop a technique for cooling a substructure of a steam generator in a nuclear reactor, wherein the technique is applicable even to a nuclear reactor that is under construction, at a test run stage, or just before replacement of a steam generator.