General Electric Company

Sampath Ranganath

John Geddes Erbes

Paul Van Diemen

Ahdee Quan Chan

Randal Raymond Stark

Eric Russel Willis

Barry Hal Koepke

Mark Olaf Lenz

Steven Max Buckner

In a method for sealing a hollow, elongate member within a reactor pressure vessel of a nuclear reactor, a section of the elongate member may be removed to separate the elongate member into an upper portion and a lower portion with an opening there between. The lower portion may be attached to a surface of the reactor pressure vessel in-situ, so as to seal off a potential leakage path through the elongate member.

Methods of repairing leaking elongate hollow members in boiling water reactors

The disclosed invention is a system and a method utilizing a web-based interactive database to automate the process for managing internal components of a plant. The system captures all essential information and provides on-line up-to-date information upon request. The system provides valuable information to a customer to analyze the plant specific problem and offers inspection methods, mitigation methods and processes to resolve the problem.

Systems and methods for managing assets using an interactive database

The method facilitates sealing a control rod drive (CRD) housing to a stub tube in a reactor pressure vessel of a nuclear reactor, if any cracks develop proximate a heat affected zone of the stub tube at a CRD housing re-attachment weld. The method includes cutting the CRD housing to separate an upper portion of the CRD housing from a lower portion of the CRD housing. The lower portion remains in place and in substantial alignment with the upper portion of the CRD housing. A lower portion member, which may be the lower portion that was cut, is re-attached to the reactor vessel by a re-attachment weld at a location below potential differential leakage paths along the CRD housing. The heat affected zone (e.g., at the re-attachment weld) is covered with a corrosion resistant material to prevent future leaking, and to protect the weld from inter-granular stress corrosion cracking (IGSCC).

Method of repairing leaking elongate hollow members in boiling water reactors

A method is provided for managing inspection requirements using a network-based system. The system includes a server system coupled to a centralized database and at least one client system. The method includes receiving information relating to a plurality of components of a specific plant and storing the information into a centralized database. The method further includes cross-referencing the information received, updating the centralized database based on the information received and providing information in response to an inquiry.

Systems and methods for managing inspections

A seal apparatus for a jet pump slip joint in a boiling water nuclear reactor pressure vessel, in an exemplary embodiment, includes a split seal ring and a segmented diaphragm spring engaging the split seal ring at an inner circumference of the diaphragm spring. The diaphragm spring includes a plurality of latch assemblies spaced circumferentially around an outer circumference. A plurality of slots, spaced circumferentially around the inner circumference, extend from the inner circumference to the support portion. Each latch assembly includes a latch bolt extending through and threadendly engaging a corresponding latch bolt opening in the diaphragm spring. Each latch bolt includes a head and a plurality of ratchet teeth spaced around the periphery of the latch bolt head. A locking spring is positioned to engage the ratchet teeth of the latch bolt head. The latch assembly further includes a latch arm coupled to the latch bolt. The latch arm includes a slot sized to receive a diffuser guide ear.

Jet pump slip joint seal

A repair apparatus for a shroud in a nuclear reactor pressure vessel is described. 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. The upper stabilizer wedge includes an integral leaf spring portion formed by a slot in the wedge and is configured to engage the side wall of the reactor pressure vessel. 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. A horizontal stabilizing spring is attached to the wedge and 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.

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San Jose, CA, United States of America

Sampath Ranganath