General Electric Company

Gerald M Gordon

Robert L Cowan, Ii

Alvin J Jacobs

James H Terhune

Harvey Donald Solomon

Robert C Dixon

James E Cearley

Paul Van Diemen

Edwin D Sayre

Beth Ann McAllister

Thomas Pompilio Diaz

Robert James Law

Ronald B Adamson

James Edward Charnley

Mickey O Marlowe

Richard A Proebstle

Richard W Perry

John H Davies

Carl P Ruiz

David Wesley Sandusky

Barry M Gordon

Donald Allan Hale

Randall N Robinson

George E Petersen

Daniel Weinstein

Gary W Contreras

William J Marble

Veronica L McCarthy

David J Coates

An improved stainless steel composition and an improved stainless Ni-based alloy are disclosed for use in nuclear reactor environments. The improved stainless alloys include a stainless steel and a stainless nickel based alloy that are particularly well adapted for use in reactor components that are positioned adjacent to components formed from a zirconium based metal. The improved stainless steel is an austenitic stainless steel material that include less than approximately 0.2 percent manganese by weight, but does include sufficient austenitic stabilizer to prevent a martensitic transformation during fabrication or use in a reactor environment. In one preferred embodiment, the stainless alloy is formed without any significant amount of manganese.

Reduction of manganese content of stainless alloys to mitigate corrosion

A passive catalytic ammonia converter operating in the water/steam mixture exiting the core of a boiling water reactor. The catalytic ammonia converter is made of catalytic material arranged and situated such that substantially all of the water/steam mixture entering the water/steam separator device flows over the surface of the catalytic material. The catalytic surfaces react ammonia and/or NO with O.sub.2 or H.sub.2 O.sub.2 in the water/steam mixture to form nitrite or nitrate. The passive catalytic ammonia converter is constructed to ensure that the pressure drop of the reactor water across the device is very small. The catalytic ammonia converter can include a plurality of stainless steel flow-through housings packed with catalytic ammonia converter material, which could take the form of tangled wire or strips, crimped ribbon, porous sintered metal composite or any other structure having a high surface area-to-volume ratio.

Internal passive catalytic ammonia converter for reduction of main steam

A method and an apparatus for measuring the degree of intergranular stress corrosion cracking protection in the area of the core shroud weldments. Electrochemical potential sensors are attached to the shroud wall and the upper core spray line at locations near the actual weld surface of concern to ensure the accuracy of the electrochemical potential measurements. A working electrode doped or coated with noble metal and a reference electrode are implanted into the reactor core shroud. A throughhole is drilled or machined into the shroud wall at an elevation just below the top guide support ring, i.e., near the core shroud weldments. Then a plug containing the working and reference electrodes is inserted into the throughhole.

Shroud electrochemical potential monitor

A catalytic recombiner device for reacting two or more molecular species having dilute concentrations in fluids flowing in pipes at elevated temperatures. The species are stable in the bulk fluid, but because of a high electrochemical potential, the species create conditions favorable to stress corrosion cracking in the pipe walls. If the pipe forms a portion of a coolant system, as in a nuclear power plant, the dissolved chemical species are transported and distributed throughout the system with undesirable consequences. To reduce the electrochemical potential, a cartridge having catalytic surfaces is installed in the flow upstream of the component to be protected against stress corrosion cracking. The catalytic surfaces of the cartridge form a small amount of benign reaction product (e.g. water), thereby reducing the concentration of undesirable species. The presence of a compact catalytic recombiner at strategic locations substantially reduces the electrochemical potential and protects both the piping and the vessels and internals connected to the piping from corrosive attack.

Method and apparatus for local protection of piping systems from stress

Electrically controlled load activating mechanisms that can be used inside the containment vessel of a nuclear reactor in conjunction with bellows-loaded DCB crack growth sensors installed inside the reactor pressure vessel or piping of a nuclear reactor. One mechanism is a liquid-filled, double-bellows master/slave arrangement connected by a capillary tube to transmit the loading provided by a linear motion device. Another mechanism uses a heat-resistant gas bottle that can be heated in a furnace to increase the gas pressure to expand the bellows of the DCB sensor. A third mechanism uses a pump or compressor to provide the necessary expansion force. The loading is controlled via electrical connections that do not require special pressure boundary penetrations of the containment vessel.

Remote load activating mechanisms

An apparatus and method for monitoring corrosion to members within the core of a nuclear reactor, particularly fuel rod cladding. A sensor means is submerged inside the core of a nuclear reactor near the member or fuel rods. The sensor means is comprised of a generally cylindrical section having an outer surface that is subject to corrosion and radiation, and has a cross-sectional area A.sub.1. The sensor means additionally has a reference section subjected to radiation but not to corrosion, and having a cross-sectional area A.sub.2. At least one pair of first probes, separated by a length L.sub.1, is placed in electrical contact with the cylindrical section. At least one pair of second probes separated by a length L.sub.2, is placed in electrical contact with the reference section. A current is passed throughout the sensor means to produce a potential gradient in the cylindrical section and reference section. The change in potential in the reference section, and the cylindrical section is measured and used to calculate the subsequent cross-sectional area in the cylindrical section between the first probes. Preferably, the cylindrical section and reference section are made from the same material as the member.

Method for monitoring corrosion on a member in a nuclear reactor core

An apparatus and method for monitoring corrosion to members within the core of a nuclear reactor, particularly fuel rod cladding. A sensor means is submerged inside the core of a nuclear reactor near the member of fuel rods. The sensor means is comprised of a generally cylindrical section having an outer surface that is subject to corrosion and radiation, and has a cross-sectional area A.sub.1. The sensor means additionally has a reference section subjected to radiation but not to corrosion, and having a cross-sectional area A.sub.2. At least one pair of first probes, separated by a length L.sub.1, is placed in electrical contact with the cylindrical section. At least one pair of second probes separated by a length L.sub.2, is placed in electrical contact with the reference section. A current is passed throughout the sensor means to produce a potential gradient in the cylindrical section and reference section. The change in potential in the reference section, and the cylindrical section is measured and used to calculate the subsequent cross-sectional area in the cylindrical section between the first probes. Preferably, the cylindrical section and reference section are made from the same material as the member.

Sensor for monitoring corrosion on a member in a nuclear reactor core

Deposition of radioactive cobalt on the interior surfaces of a water-cooled nuclear reactor and intergranular stress corrosion cracking are inhibited or substantially prevented by the continuous injection of zinc oxide to the reactor water. The zinc oxide may be prepared in the form of a paste, a slurry, or a preformed aqueous solution.

Inhibition of radioactive cobalt deposition in water-cooled nuclear

An apparatus and process for control rod assembly and construction for a nuclear reactor is disclosed wherein the control rod is of cruciform configuration having a plurality of and preferably four flat planar members. Each planar member is made up with an assembly of square cross sectioned tubular members. Each square cross sectioned tubular member includes a defined cylindrical center and a constant thickness surrounding tubular metallic shell to provide the contained neutron absorbing materials in a cylindrically shaped pressure vessel. The constant thickness surrounding tubular shell constituting the cylindrical shaped pressure vessel has four corner sections added to generate a modified square exterior profile. These added four corner sections comprise an addition to the tubular side wall thickness of the shell. Appropriate chamfers and rounding are provided to the square profile to impart both stress relief and convenient points for automated welding between side-by-side square sections. The square cross sectioned tubular members when welded together extend axially and longitudinally the length of the control rod in the direction of control rod movement and fluid flow. The square cross sectioned tubular members are filled with a neutron absorbing material, plugged at both ends, and fastened by welding in side-by-side relation to form the planar control rod members of the required cruciform shape. These square cross sectioned tubular members as welded in side-by-side relation are in integral portion of the longitudinal support structure of the control rod and fasten and interconnect the handle at the top of the control rod and a conventional velocity limiting device at the bottom of the control rod. The fastening of the handles and velocity limiting device at the respective top and bottom provides a secondary and supplemental seal to primary plugs containing the neutron absorbing material. If needed for enhanced life, provision is made for the incorporation of a hafnium plate in a rectangular sectioned tube adjacent the handle at upper end of the control rod, this tip being variable in length dependent upon individual control rod requirements. Thus square sectioned tubular members form an improved poison containing cylindrical volume for providing resultant improved control rod worth and increased in service life time in moderating and shutting down nuclear reactions. Consequently, the disclosed construction can be used both for conventional control rods as well as for control rod configured for control cell locations.

Control rod

An apparatus and process for control rod assembly and construction for a nuclear reactor is disclosed wherein the control rod is of cruciform configuration having a plurality of and preferably four flat planar members. Each planar member is made up with an assembly of square cross sectioned tubular members. Each square cross sectioned tubular member includes a defined cylindrical center and a constant thickness surrounding tubular metallic shell to provide the contained neutron absorbing materials in a cylindrically shaped pressure vessel. The constant thickness surrounding tubular shell constituting the cylindrical shaped pressure vessel has four corner sections added to generate a modified square exterior profile. These added four corner sections comprise an addition to the tubular side wall thickness of the shell. Appropriate chamfers and rounding are provided to the square profile to impart both stress relief and convenient points for automated welding between side-by-side square sections. The square cross sectioned tubular members when welded together extend axially and longitudinally the length of the control rod in the direction of control rod movement and fluid flow. The square cross sectioned tubular members are filled with a neutron absorbing material, plugged at both ends, and fastened by welding in side-by-side relation to form the planar control rod members of the required cruciform shape. These square cross sectioned tubular members as welded in side-by-side relation are an integral portion of the longitudinal support structure of the control rod and fasten and interconnect the handle at the top of the control rod and a conventional velocity limiting device at the bottom of the control rod. The fastening of the handles and velocity limiting device at the respective top and bottom provides a secondary and supplemental seal to primary plugs containing the neutron absorbing material. If needed for enhanced life, provision is made for the incorporation of a hafnium plate in a rectangular sectioned tube adjacent the handle at upper end of the control rod, this tip being variable in length dependent upon individual control rod requirements. Thus square sectioned tubular members form an improved poison containing cylindrical volume for providing resultant improved control rod worth and increased in service life time in moderating and shutting down nuclear reactions. Consequently, the disclosed construction can be used both for conventional control rods as well as for control rod configured for control cell locations.

A heat treatment method for inhibiting irradiation induced stress corrosion cracking in stainless steel and related nickel-chromium alloys.

Treatment for inhibiting irradiation induced stress corrosion cracking

A chromium-nickel austenitic stainless steel alloy composition, including specific proportions of carbon with a combination of niobium and tantalum.

Austenitic stainless steel alloy

In a control rod for a nuclear reactor having a cruciform shape with four flat planar members, an improved membrane for safeguarding against crevice cracking corrosion is disclosed. Each of the four flat planar members of the control rod is fabricated from side-by-side tubular members containing sealed neutron absorbing poisons. Each of the tubular members has square outside sections. These square outside sections are welded together to form the flat planar members of the control rod. The improvement includes a protective membrane wrapped over the structural flat planar member. The protective membrane surrounds the member and is maintained to the planar member under compression by ambient reactor pressure. The membrane provides an additional margin against crevice corrosion cracking by preventing the water of the reactor from coming into contact with the welded side-by-side square sectioned tubular members. Because the membrane is not under tension, it is not subject to either irradiation assisted stress corrosion cracking nor conventional stress corrosion cracking.

BWR control rod

Stress corrosion cracking of austenitic stainless steel or nickel-based alloys attributable at least in part to exposure to irradiation is reduced with a specific heat treatment.

Treatment for overcoming irradiation induced stress corrosion cracking

A sensor apparatus cooperates with an electrical system capable of delivering power to and receiving signals from the sensor. A cable housing insulated wires is connected to the sensor and disposed in proximity to a fluid possibly containing a corrosive impurity. Corrosion detecting elements are connected to the wires and are immersed in the fluid possibly containing the impurity. Further, the elements are provided with breakable regions fabricated to corrode in the presence of the corrosive impurity. To facilitate fracture, the elements are loaded in tension so the breakable regions will fracture rapidly in the presence of corrosive impurities, to thereby send an electric signal to remote electrical equipment to indicate the presence of at least one impurity.

Growing community of inventors

Soquel, CA, United States of America

Gerald M Gordon