General Electric Company

Kenneth R Izzo

James E Cearley

James Edward Charnley

Robert C Dixon

Charles A Dalke

David L Faulstich

Dane T Snyder

Hsueh-Wen Na Pao

Louis L Aiello

Robert K Lovell

Johnny T Ma

Martin Sullivan Laurent

Charles E Burdg

Tsi-Pin Choong

Don G Umble

Joel C Swanson

Peter F Kachel

Terry K Dallas

Gary H Hankin

An unlatching tool which can be used to uncouple and grapple standard handle and extended handle control rods from their associated control rod drives. The unlatching tool is lowered onto the handle of a control rod residing in the reactor core. Due to differences in the shapes of the handles of the standard handle and extended handle control rods, the tool is able to sense the type of control rod present and, through the use of a pneumatically actuated cylinder, position the unlatching mechanism at the appropriate elevation. In particular, a four-way valve, that is actuated by contact with the top corner of the handle of the extended handle control rod and is not contacted by the handle of the standard handle control rod, is used to differentiate between control rod types. The unlatching mechanism is positioned relative to the unlocking handle on the control rod using a pneumatic cylinder activated by the four-way valve. The cylinder has a stroke length equal to the difference in length between the standard handle and extended handle control rods.

Tool for unlatching extended handle and standard handle control rods

A process and a system for detecting defective nuclear fuel assemblies in situ. The system includes two basic subsystems. The first subsystem is a sample collection system consisting of a hood placed over the fuel. A technique whereby a vacuum is drawn across a submerged nozzle of a degas tank is used to draw up the sample of fluid out of a fuel assembly. The second subsystem determines the amount of Kr-85 radioisotope contained in the fluid sample. This information is used to determine whether the fuel assembly contains a defective, i.e., leaking, fuel rod. The measurement values for Kr-85 in the fluid sample extracted from the fuel assembly are compared to the measurement values for Kr-85 in fluid samples taken from the reactor pool prior to fuel assembly testing and fluid samples taken from non-leaking control cells, the latter values representing the background to which the subject fuel assembly is compared. Based on this comparison, a determination is made concerning whether the subject fuel assembly is defective.

Method and apparatus for in situ detection of defective nuclear fuel

In a control rod construction for moderating and controlling a reaction of a nuclear reactor, a control rod structure is disclosed in which hafnium is welded to stainless steel. Typically, the control rod is confined for movement axially along its length into and out of the reactor. The hafnium is surrounded in the control rod structure by a sheath of stainless steel to inhibit movement relative to the control rod structure other than in the axial direction of control rod movement. The hafnium is friction welded to the stainless steel of the control rod so as to stress the weld in either tension or compression due to dynamic loading occurring from control rod movement. The stainless steel to which the hafnium is welded is in turn conventionally welded or mechanically connected to the remaining steel components of the control rod. There results a secure control rod structure having welded hafnium that can withstand the dynamic loadings of fast insertion of control rods such as those loadings which commonly occur under rapid control rod acceleration and deceleration due to SCRAM of a reactor.

Hafnium stainless steel absorber rod for control rod

A control rod for a nuclear reactor wherein plates of neutron absorbing material (e.g. hafnium) are used in the upper portion thereof and tubes containing a different neutron absorbing material (e.g. boron) are positioned beneath the plates. To relieve the tubes of the weight of the plates, the plates are attached to support members of the control rod frame, to prevent or reduce the area of a gap between the plates and the tubes, joints therebetween are provided.

Nuclear reactor control rod

A microprocessor controls pressure transducer testing, calibration, and adjustment. A transducer or unit under test (UUT) receives a fluid pressure signal from a test line, and outputs a corresponding electrical signal. A test pressure is output from a fluid pressure module which is connected to the UUT. A human operator is interrogated by the microprocessor based control unit as to (1) identifying details about the desired pressure UUT to be tested, and (2) the anticipated electrical range to be encountered in reply to an applied test pressure. The microprocessor then sequentially generates electrical output signals which are sent to the test lines to generate the test pressure pulses from a fluid pressure supply. The pressure test pulses are delivered to the UUT. As the UUT receives the sequence of pressure test pulses, the microprocessor records the output electrical signals generated by the UUT, and compares them to an internal reference pressure transducer output. Display of the results occurs with preferably the percent error calculated at each step. Provision for pressure transducer UUT calibration is made by providing the calibration instrument with the ability to issue tonal and visual signals in response to UUT adjustments, thus providing sense and amount of adjustments of the zero point and span range of the transducer under test.

Instrument for calibrating pressure gauges

Growing community of inventors

San Jose, CA, United States of America

Kenneth R Izzo