Induced nuclear reactions: processes – systems – and elements – Testing – sensing – measuring – or detecting a fission reactor... – Leak detection
Reexamination Certificate
1999-06-03
2002-02-05
Carone, Michael J. (Department: 3641)
Induced nuclear reactions: processes, systems, and elements
Testing, sensing, measuring, or detecting a fission reactor...
Leak detection
C376S251000
Reexamination Certificate
active
06345082
ABSTRACT:
FIELD OF THE INVENTION
The present invention relates to nuclear fuel assemblies and, in particular, to an apparatus for detecting the failure of nuclear fuel in a nuclear fuel assembly.
BACKGROUND OF THE INVENTION
As a nuclear reactor operates and generates power, the nuclear fuel is gradually consumed and it becomes necessary at periodic intervals to inspect for failure of nuclear fuel rods composing the nuclear fuel assemblies. Such failures include a breach of the cladding of the fuel rod permitting the escape of fission products such as radioactive iodine, xenon and krypton into the reactor coolant water which circulates through the reactor core. In commercial nuclear reactors, the core comprises nuclear fuel assemblies consisting of nuclear fuel rods. The fuel rods comprise a circular or cylindrical housing commonly known as the cladding within which are stacked nuclear fuel pellets leaving a plenum space above the fuel columns and which are sealed at both ends. Failure of the cladding could result in contamination of the coolant by the escape of radioactive products from the fuel rods and which could interfere with plant operations. In practice, leak detection is not normally carried out on individual fuel rods but on fuel assemblies containing several fuel rods. Leak detection of a fuel assembly takes place by measuring fission products in a gas and/or water sample which is taken from a fuel assembly and is commonly referred to as “fuel sipping”.
Some methods involve isolating a fuel assembly in a test chamber filled with water. This has the disadvantage in that the fuel assembly must be removed from the reactor core and placed within the chamber which is time consuming, particularly when conventional nuclear reactors contain several hundred nuclear fuel assemblies in the reactor core. In other methods, sipping tests are performed on nuclear fuel assemblies while they are still positioned inside the reactor core, eliminating the need for time consuming fuel assembly movements in accomplishing the testing of the fuel. The intent of the sipping tests is to detect assemblies that contain failed fuel rods so that these assemblies can be removed from the reactor and further examined or repaired. In accomplishing these sipping tests, the reactor head and upper internals are removed, thus exposing the tops of the fuel assemblies in the core. The reactor vessel is water filled, and a continuous water coolant flow is maintained to remove decay heat from the fuel assemblies. The basic principle of the in-core sipping technique typically involves (1) restricting the coolant flow by the application of air pressure within a hood overlying the fuel assembly resulting in a temperature increase in the fuel assembly; then (2) sampling an air bubble trapped above the fuel assembly(ies) by the hood fit over the fuel assembly for gaseous fission product activity; and (3) sample the water in the fuel assembly. The temperature increase results in an internal pressure increase leading to the release of the radioactive fission products from the interior of a failed rod through the rod defect. By measuring for the presence and quantity of the radioactive isotopes in the collected gas and/or water samples taken from the assembly, the assembly can be identified as containing one or more failed fuel rods.
Thus, current techniques employed for in-core sipping depend upon achieving a fuel assembly temperature rise to release the radioactive fission products from the failed fuel rods.
However, at higher reactor shutdown coolant flow rates, the necessary temperature rise is difficult to achieve.
In order to successfully detect which assemblies have failed, it is most advantageous to selectively be able to increase the temperature of the fuel assemblies being tested, to temperatures above their normal reactor shutdown temperature. To accomplish this, the normal reactor shutdown flow rate through the fuel assemblies must be reduced or stopped during testing. The standard techniques employed in existing systems to reduce flow rates through the fuel assembly being tested involve either creating a pressurized air bubble within the test hood placed over the top of the fuel assembly or involve effectively increasing the column length of water within each fuel assembly by extending the fuel channel heights within the hood. These techniques merely add flow “resistance” to the normal reactor coolant flow in the tested channels.
The techniques currently employed to restrict flow are only partially effective. As a result, at higher levels of reactor coolant flow, the current techniques cannot restrict flow enough to permit the necessary fuel rod temperature rise to occur. This results in extended test times and/or inaccurate and unreliable test results.
It would therefore be an advantage over prior art designs to provide an apparatus for detecting failed fuel elements from a BWR, that provides an effective way to restrict coolant flow, and to thereby effect a temperature rise of the fuel rods in the fuel assembly being tested over a wide range of reactor shutdown coolant flow conditions.
SUMMARY OF THE INVENTION
In accordance with one embodiment of the present invention, a fuel assembly mechanical flow restriction apparatus is provided for detecting failure of a nuclear fuel rod in a nuclear fuel assembly situated in a reactor core of a boiling water reactor, the reactor core comprising a plurality of nuclear fuel assemblies comprising parallel fuel rods supported at an upper end by an upper tie plate and an outer channel surrounding the fuel rods for the passage of reactor coolant from a lower end to the upper end of the fuel assembly, the outer channel having upper edges, the upper end of the fuel assembly passing through and being supported by a reactor core top guide structure, the fuel assembly mechanical flow restriction apparatus comprising a testing hood comprising a top plate and side plates to form a structure with an open bottom forming an internal volume for positioning over the tops of at least one of the nuclear fuel assemblies and for receiving gases escaping from a failed fuel rod within the fuel assembly, the side plates for resting on the reactor core top guide structure, and a flow restrictor positioned within the testing hood and over at least one of the nuclear fuel assemblies, the flow restrictor comprising a sealing plate for positioning on the upper edges of the outer channel of the fuel assembly for mechanically blocking fuel assembly coolant flow exiting the upper end of the fuel assembly, and a probe assembly having at least one probe head for sampling the coolant water within the fuel assembly for detecting failure of a nuclear fuel rod in the nuclear fuel assembly, and means for causing the sealing plate of the flow restrictor to be positioned on the upper edges of the outer channel of the fuel assembly for mechanically blocking fuel assembly coolant flow from exiting the upper end of the fuel assembly and for causing the probe head to be immersed in the fuel assembly reactor coolant water within the outer channel.
In accordance with another embodiment of the present invention, a fuel assembly mechanical flow restriction apparatus is provided for detecting failure of a nuclear fuel rod in a nuclear fuel assembly situated in a reactor core of a boiling water reactor, the reactor core comprising a plurality of nuclear fuel assemblies comprising parallel fuel rods supported at an upper end by an upper tie plate and an outer channel surrounding the fuel rods for the passage of reactor coolant from a lower end to the upper end of the fuel assembly, the outer channel having upper edges, the upper end of the fuel assembly passing through and being supported by a reactor core top guide structure, the fuel assembly mechanical flow restriction apparatus comprising a testing hood comprising a top plate and side plates to form a structure with an open bottom forming an internal volume for positioning over the tops of at least one of the nuclear fuel assemblies and for receiving gases escaping fr
Galioto Thomas A.
Lee Jerome B.
McClelland Richard G.
O'Leary Patrick M.
Ruzauskas Edward J.
Carone Michael J.
Kenyon & Kenyon
Richardson John
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