Patent Number: 056028881
Section: summary

FIELD OF THE INVENTION This invention relates to reducing the corrosion potential of components exposed to high-temperature water. As used herein, the term "high-temperature water" means water having a temperature of about 150.degree. C. or greater or steam. High-temperature water can be found in a variety of known apparatus, such as water deaerators, nuclear reactors, and steam-driven power plants. 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 288.degree. C. for a boiling water reactor (BWR), and about 15 MPa and 320.degree. C. 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, and nickel-based, cobalt-based and zirconium-based alloys. Despite careful selection and treatment of these materials for use in water reactors, corrosion occurs on 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, crevice geometry, 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 1 to 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 kinetic 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.sub.2, H.sub.2 O.sub.2, O.sub.2 and oxidizing and reducing radicals. For steady-state operating conditions, equilibrium concentrations of O.sub.2, H.sub.2 O.sub.2, and H.sub.2 are established in both the water which is recirculated and the steam going to the turbine. This concentration of O.sub.2, H.sub.2 O.sub.2, and H.sub.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 dissolved hydrogen to the reactor feedwater. When the hydrogen reaches the reactor vessel, it reacts with the radiolytically formed oxidizing species on metal surfaces 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 -230 to -300 mV 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 present results in an ECP below the critical potential. It has been shown that IGSCC of Type 304 stainless steel (containing 18-20% Cr, 8-10.5% Ni, 2% Mn, remainder Fe) used in BWRs can be mitigated by reducing the ECP of the stainless steel to values below -230 mV(SHE). An effective method of achieving this objective is to use HWC. However, high hydrogen additions, e.g., of about 200 ppb or greater, that may be required to reduce the ECP below the critical potential, can result in a higher radiation level in the steam-driven turbine section from incorporation of the short-lived N-16 species in the steam. For most BWRs, the amount of hydrogen addition required to provide mitigation of IGSCC of pressure vessel internal components results in an increase in the main steam line radiation monitor by a factor of five to eight. This increase in main steam line radiation can cause high, even unacceptable, environmental dose rates that can require expensive investments in shielding and radiation exposure control. Thus, recent investigations have focused on using minimum levels of hydrogen to achieve the benefits of HWC with minimum increase in the main steam radiation dose rates. An effective approach to achieve this goal is to either coat or alloy the alloy surface with palladium or other noble metal. Palladium doping has been shown to be effective in mitigating the crack growth rate in Type 304 stainless steel, Alloy 182 having the composition in wt.%: Ni, 59.0 min.; Cr, 13.0-17.0; Fe, 10.0 max.; Mn, 5.0-9.5; Si, 1.0 max.; Cu, 0.5 max.; Ti, 1.0 max.; S, 0.015 max.; C, 0.10 max.; P, 0.03 max.; (Nb+Ta), 1.0-2.5; other, 0.5 max., and Alloy 600 having the nominal composition in wt.%: Cr, 16.0; Fe, 8.0; Si, 0.5; Cu, 0.5 max.; Ti, 0.3 max.; C, 0.08; Ni, balance. The techniques used to date for palladium coating include electroplating, electroless plating, hyper-velocity oxy-fuel, plasma deposition and related high-vacuum techniques. Palladium alloying has been carried out using standard alloy preparation techniques. These approaches are ex-situ techniques in that they cannot be practiced while the reactor is in operation. Also noble metal coatings such as those applied by plasma spraying and by hyper-velocity oxy-fuel must be applied to all surfaces that require protection, i.e., they afford no protection to adjacent uncoated regions. The most critical requirement for IGSCC protection of Type 304 stainless steel is to lower its ECP to values below the protection potential, i.e., -230 mV(SHE). The manner in which this potential is achieved is immaterial, e.g., by alloying, doping or by any other method. It has been demonstrated that it is sufficient to dope the oxide film by the appropriate material (e.g., Pd) to achieve a state of lower ECP in the presence of low levels of hydrogen. It was shown in later work that a thickness of 200-300 .ANG. of the doping element (Pd) is sufficient to impart this benefit of lower potential. This is not surprising because the ECP is an interfacial property, and hence modifying the interface by a process such as doping would alter its ECP. The critical requirement is that the dopant remain on the surface over a long period of time to gain the maximum benefit from the doping action. One method of in-situ application of a noble metal onto stainless steel or other metal surfaces inside a boiling water reactor is by injecting a decomposable noble metal compound into the high-temperature (i.e., 550.degree. F.) water that is in contact with the metal surface during reactor operation. As a result of decomposition of the noble metal compound, the oxide film on the metal surfaces becomes doped with noble metal. The amount of noble metal dopant can be made high enough to provide sufficient catalytic activity for H.sub.2 and O.sub.2 recombination to reduce the ECP of the metal surfaces to required protection values. This approach of noble metal doping has been shown to be effective against crack initiation and crack growth in stainless steel at H.sub.2 /O.sub.2 molar ratios greater than 2 in the reactor environment. SUMMARY OF THE INVENTION The present invention is an alternative method for the application of palladium or other noble metal onto stainless steel or other metal surfaces inside a boiling water reactor. The method comprises the step of injecting a solution of a compound containing a noble metal into the coolant water while the reactor is shutdown or during heatup with only recirculation pump heat, i.e., without nuclear heat generation. As used herein, the term "solution" refers to both solutions and suspensions of a noble metal compound. During shutdown, the reactor coolant water reaches temperatures as low as 120.degree. F., in contrast to the water temperature of 550.degree. F. during normal operation. On the other hand, pump heat can bring the water temperature up to 400.degree.-450.degree. F. At these reduced temperatures, the rate of thermal decomposition of the injected noble metal compound is reduced. However, decomposition of the noble metal compound is also induced by radiation produced inside the reactor. In particular, the noble metal compound can be decomposed by the gamma radiation emanating from the nuclear fuel core of the reactor. Decomposition can also be precipitated by radiation from radioactive isotopes in the material of the reactor component to be protected by noble metal doping. Thus the noble metal compound decomposes under reactor thermal and radiation conditions to release atoms of the noble metal which incorporate in or deposit on the oxide film formed on the stainless steel and other alloy components. As used herein, the term "atoms" also includes ions of the noble metal. The preferred compound for use in noble metal doping is palladium acetylacetonate. The concentration of palladium in the reactor water is preferably in the range of 5 to 100 ppb. Upon injection, the palladium acetylacetonate decomposes and deposits palladium on the crudded (heavily oxidized) stainless steel and other alloy surfaces immersed in the water. The palladium is incorporated into the oxide film or crud via a process wherein palladium ions/atoms apparently replace iron, nickel and/or chromium atoms in the oxide film or crud, resulting in palladium doping. Alternatively, palladium may be deposited within or on the surface of the oxide film or crud in the form of a finely divided metal. The oxide film is believed to include mixed nickel, iron and chromium oxides. The passive oxide films on the surfaces of structural materials can be doped or coated with palladium or other noble metal using either in situ or ex situ techniques. In accordance with both techniques, the structural material is immersed in a solution or suspension of a compound containing the noble metal. Decomposition of the noble metal compound in the solution is then induced by exposing the immersed material to electromagnetic radiation, e.g., ultraviolet or gamma radiation. Even at the low temperatures reached during reactor shutdown or during heatup with recirculation pump heat, the radiation-induced decomposition in combination with thermal decomposition provide noble metal doping of oxide film or crud sufficient to reduce the ECP at the oxide film/water interface to below the critical threshold potential, thereby mitigating stress corrosion cracking.