Induced nuclear reactions: processes – systems – and elements – Handling of fission reactor component structure within...
Reexamination Certificate
2000-12-12
2002-10-29
Jordan, Charles T. (Department: 3641)
Induced nuclear reactions: processes, systems, and elements
Handling of fission reactor component structure within...
C376S305000, C376S306000, C702S023000
Reexamination Certificate
active
06473480
ABSTRACT:
TECHNICAL FIELD OF THE INVENTION
This invention relates generally to reducing the corrosion potential of components exposed to high-temperature water through a noble metal application process. More specifically, the invention relates to a method and apparatus for modeling and maintaining the amount of noble metals deposited in the water circuit of a boiling water reactor and components thereof during an in situ noble metal application process.
BACKGROUND OF THE INVENTION
Nuclear reactors are used in electric power generation, research and propulsion. A reactor pressure vessel contains the reactor coolant, i.e. water, which removes heat from the nuclear core. Respective piping circuits carry the heated water or steam to the steam generators or turbines and carry circulated water or feedwater back to the vessel. Operating pressures and temperatures for the reactor pressure vessel are about 7 MPa and 288EC for a boiling water reactor (BWR), and about 15 MPa and 320EC for a pressurized water reactor (PWR). The materials used in both BWRs and PWRs must withstand various loading, environmental and radiation conditions.
Some of the materials exposed to high-temperature water include carbon steel, alloy steel, stainless steel, nickel-based, cobalt-based and zirconium-based alloys. Despite careful selection and treatment of these materials for use in water reactors, corrosion occurs in the materials exposed to the high-temperature water. Such corrosion contributes to a variety of problems, e.g., stress corrosion cracking, crevice corrosion, erosion corrosion, sticking of pressure relief valves and buildup of the gamma radiation-emitting Co-60 isotope.
Stress corrosion cracking (SCC) is a known phenomenon occurring in reactor components, such as structural members, piping, fasteners and welds exposed to high-temperature water. As used herein, SCC refers to cracking propagated by static or dynamic tensile stressing in combination with corrosion at the crack tip. The reactor components are subject to a variety of stresses associated with, e.g., differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other sources such as residual stress from welding, cold working and other asymmetric metal treatments. In addition, water chemistry, welding, heat treatment, and radiation can increase the susceptibility of metal in a component to SCC.
It is well known that SCC occurs at higher rates when oxygen is present in the reactor water in concentrations of about 5 ppb or greater. SCC is further increased in a high radiation flux where oxidizing species, such as oxygen, hydrogen peroxide, and short-lived radicals, are produced from radiolytic decomposition of the reactor water. Such oxidizing species increase the electrochemical corrosion potential (ECP) of metals. Electrochemical corrosion is caused by a flow of electrons from anodic to cathodic areas on metallic surfaces. The ECP is a measure of the thermodynamic tendency for corrosion phenomena to occur, and is a fundamental parameter in determining rates of, e.g., SCC, corrosion fatigue, corrosion film thickening, and general corrosion.
In a BWR, the radiolysis of the primary water coolant in the reactor core causes the net decomposition of a small fraction of the water to the chemical products H
2
, H
2
O
2
, O
2
and oxidizing and reducing radicals. For steady-state operating conditions, equilibrium concentrations of O
2
, H
2
O
2
, and H
2
are established in both the water which is recirculated and the steam going to the turbine. This concentration of O
2
, H
2
O
2
, and H
2
is oxidizing and results in conditions that can promote intergranular stress corrosion cracking (IGSCC) of susceptible materials of construction. One method employed to mitigate IGSCC of susceptible material is the application of hydrogen water chemistry (HWC), whereby the oxidizing nature of the BWR environment is modified to a more reducing condition. This effect is achieved by adding hydrogen gas to the reactor feedwater. When the hydrogen reaches the reactor vessel, it reacts with the radiolytically formed oxidizing species to reform water, thereby lowering the concentration of dissolved oxidizing species in the water in the vicinity of metal surfaces. The rate of these recombination reactions is dependent on local radiation fields, water flow rates and other variables.
The injected hydrogen reduces the level of oxidizing species in the water, such as dissolved oxygen, and as a result lowers the ECP of metals in the water. However, factors such as variations in water flow rates and the time or intensity of exposure to neutron or gamma radiation result in the production of oxidizing species at different levels in different reactors. Thus, varying amounts of hydrogen have been required to reduce the level of oxidizing species sufficiently to maintain the ECP below a critical potential required for protection from IGSCC in high-temperature water. As used herein, the term “critical potential” means a corrosion potential at or below a range of values of about −0.230 to −0.300 V based on the standard hydrogen electrode (SHE) scale. IGSCC proceeds at an accelerated rate in systems in which the ECP is above the critical potential, and at a substantially lower or zero rate in systems in which the ECP is below the critical potential. Water containing oxidizing species such as oxygen increases the ECP of metals exposed to the water above the critical potential, whereas water with little or no oxidizing species presents results in an ECP below the critical potential.
Corrosion potentials of stainless steels in contact with reactor water containing oxidizing species can be reduced below the critical potential by injection of hydrogen into the water so that the dissolved hydrogen concentration is about 50 to 100 ppb or greater. For adequate feedwater hydrogen addition rates, conditions necessary to inhibit IGSCC can be established in certain locations of the reactor. Different locations in the reactor system require different levels of hydrogen addition. For example, much higher hydrogen injection levels are necessary to reduce the ECP within the high radiation flux of the reactor core, or when oxidizing cationic impurities, e.g., cupric ion, are present.
An effective step toward to achieving the goal of reducing ECP within the high radiation flux of the reactor core is to either coat or alloy the stainless steel surface with palladium or any other noble group metal. As used herein, the term “noble metal” means metals from the group consisting of platinum, palladium, osmium, ruthenium, iridium, rhodium, and mixtures thereof. The presence of palladium or other noble metal on the stainless steel surface catalyzes the recombination of oxidizing and reducing species in contact with the surface and reduces the injected hydrogen demand in achieving the required IGSCC critical potential of −0.230 V(SHE). Known techniques for palladium coating include electroplating, electroless plating, plasma deposition and related high-vacuum techniques. Palladium alloying can also be carried out using standard alloy preparation techniques. Unfortunately, both of these approaches are ex situ techniques in that they cannot be practiced while the reactor is in operation.
U.S. Pat. No. 5,135,709 to Andresen et al. discloses a method for lowering the ECP on components formed from carbon steel, alloy steel, stainless steel, nickel-based alloys or cobalt-based alloys which are exposed to high-temperature water by forming the component to have a catalytic layer of a platinum group metal. As used therein, the term “catalytic layer” means a coating on a substrate, or a solute in an alloy formed into the substrate, the coating or solute being sufficient to catalyze the recombination of oxidizing and reducing species at the surface of the substrate.
In nuclear reactors, ECP is increased by the high levels of oxidizing species, e.g., up to 200 ppb or greater of oxygen in the water measured in the circulation piping, produced from the ra
Kruger Richard M.
Law Robert J.
Jordan Charles T.
Matz Daniel
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