Induced nuclear reactions: processes – systems – and elements – Reactor protection or damage prevention – Core restraint means
Reexamination Certificate
2000-02-01
2002-01-29
Behrend, Harvey E. (Department: 3641)
Induced nuclear reactions: processes, systems, and elements
Reactor protection or damage prevention
Core restraint means
C376S260000
Reexamination Certificate
active
06343107
ABSTRACT:
BACKGROUND OF THE INVENTION
This invention relates generally to maintenance and repair of nuclear reactors, and more particularly, to the repair of the fuel core shroud of a boiling water nuclear reactor.
A reactor pressure vessel (RPV) of a boiling water reactor (BWR) typically has a generally cylindrical shape and is closed at both ends, e.g., by a bottom head and a removable top head. A top guide, sometimes referred to as a grid is spaced above a core plate within the RPV. A core shroud, or shroud, surrounds the core plate and is supported by a shroud support structure. The core shroud is a reactor coolant flow partition and structural support for the core components. Particularly, the shroud has a generally cylindrical shape and surrounds both the core plate and the top guide. The top guide includes a plurality of openings, and fuel bundles are inserted through the openings and are supported by the core plate.
The shroud, due to its large size, is formed by welding a plurality of stainless steel cylindrical sections together. Specifically, respective ends of adjacent shroud sections are joined with a circumferential weld. During operation of the reactor, the circumferential weld joints may experience stress corrosion cracking (SCC) in the weld heat affected zones which can diminish the structural integrity of the shroud. In particular, lateral seismic/dynamic loading could cause relative displacements at cracked weld locations, which could produce large core flow leakage and misalignment of the core that could prevent control rod insertion and a safe shutdown.
The shroud, containing welds which may experience SCC, is located in a remote, confined location below
60
feet of water and is accessible only during refueling outages. Because the loss of power production during outages is a significant cost, it is desirable to minimize the required duration of any repair operations, particularly, shroud weld repair operations.
BRIEF SUMMARY OF THE INVENTION
A repair apparatus for a shroud in a nuclear reactor pressure vessel that does not require any installation machining of existing reactor components and is quickly installed in the reactor pressure vessel. In an exemplary embodiment, the repair apparatus includes an upper stabilizer assembly, a lower stabilizer assembly, and a tie rod configured to extend between and to couple to the upper and lower stabilizer assemblies.
The upper stabilizer includes a stabilizer block and an upper stabilizer wedge slidably coupled to the upper stabilizer block. The upper stabilizer block is configured to couple to a shroud lug pair. The upper stabilizer assembly further includes a jack bolt extending through a jack bolt opening in the upper stabilizer wedge and threadedly engaging a jack bolt opening in the upper stabilizer block. The upper stabilizer wedge includes a ratchet lock spring configured to engage the jack bolt to maintain the tightness of the jack bolt. The upper stabilizer wedge further includes an integral leaf spring portion formed by a slot in the wedge and configured to engage the side wall of the reactor pressure vessel. The leaf spring portion provides flexibility for tightening the jack bolt at assembly and absorbing operating variations in the annulus width, while also limiting radial and friction interaction loads for various reactor operating conditions.
The lower stabilizer assembly includes a stabilizer block and a lower stabilizer wedge slidably coupled to the lower stabilizer block. The lower stabilizer block is configured to engage the shroud. The lower stabilizer assembly also includes a jack bolt extending through a jack bolt opening in the lower stabilizer wedge and threadedly engaging a jack bolt opening in the lower stabilizer block. Also, the lower stabilizer wedge includes a ratchet lock spring configured to engage the jack bolt to maintain the tightness of the jack bolt. A horizontal stabilizing spring is attached to the surface of the wedge that engages the reactor pressure vessel side wall. The horizontal stabilizing spring is configured to engage the side wall of the reactor pressure vessel.
The tie rod is threaded at each end. One end threadedly engages a tie rod opening in the bottom stabilizer block. The other end is received by the upper stabilizer block and is secured by a tie rod nut. The tie rod nut reacts the tie rod load against the upper stabilizer block. In one embodiment, the tie rod is fabricated from Ni—Cr—Fe alloy X-750 steel. Tie rod preload increases at operating temperatures due to the differential expansion between the X-750 tie rod and the stainless steel shroud. With an X-750 tie rod, more thermal differential contraction of the tie rod is produced than needed for the desired operating preload. To compensate, a belleville spring washer is positioned between the tie rod nut and the upper stabilizer block. The spring washer deflects only slightly with the low mechanical installation preload, for example, 5000 pounds, but compresses additionally to seat flat against the upper stabilizer block under fill thermal preload.
A limit stop is attached near the upper end of the tie rod. The limit stop includes two shear pins which fit mating holes in the bottom of the upper stabilizer block, providing a torque restraint for tightening the tie rod nut as well as a pinned anti-vibration connection to support the tie rod during operation.
The outer surface of the tie rod includes a plurality of longitudinal grooves spaced around the tie rod periphery to limit flow induced vibration of the tie rod. The grooves reduce the vortex shedding frequency below the natural vibration frequency of the tie rod, so resonant excitation of the tie rod does not occur.
The above described shroud repair apparatus is quickly and easily installed in a reactor pressure vessel because it does not require any installation machining of existing reactor components. The lower stabilizer assembly and tie rod are pre-assembled with the tie rod threaded into the lower stabilizer block. This assembly is lowered into position in the annulus with the lower stabilizer engaging the protruding core plate support ledge. The lower stabilizer wedge is then lowered into place on the lower stabilizer block and adjusted by tightening the jack bolt. The ratchet lock spring prevents the jack bolt from loosening.
The upper stabilizer assembly is lowered into position in the annulus area between the shroud and the reactor pressure vessel outer wall, engaging the tie rod through the center hole in the upper stabilizer block. A lug opening at the top of the upper stabilizer block is then engaged onto a shroud lug pair. The tie rod nut is then lowered in place and tightened to the tie rod which causes the lower stabilizer block to seat against the bottom surface of the shroud core plate support ledge. A ratchet lock spring prevents the tie rod nut from loosening during reactor operation. The upper stabilizer wedge is then lowered into position and adjusted by tightening the jack bolt. The ratchet lock spring prevents the jack bolt from loosening during reactor operation. Typically four repair apparatus, equally spaced around the shroud, are installed in a reactor pressure vessel to repair cracked shroud welds.
The above described shroud repair apparatus does not require any installation machining of existing reactor components prior to installation, and therefore is quickly and easily installed in the reactor pressure vessel. The repair apparatus provides lateral support for the shroud and imparts a clamping force to the shroud to maintain shroud joint integrity and overcome the effects of any stress corrosion cracking in the circumferential shroud welds.
REFERENCES:
patent: 5402570 (1995-04-01), Weems et al.
patent: 5538381 (1996-07-01), Erbes
patent: 5621778 (1997-04-01), Erbes
patent: 5675619 (1997-10-01), Erbes et al.
patent: 5742653 (1998-04-01), Erbes et al.
Erbes John G.
Koepke Barry H.
Ranganath Sampath
Armstrong Teasdale LLP
Behrend Harvey E.
General Electric Company
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