MODIFICATION OF OXIDE FILM ELECTRICAL CONDUCTIVITY TO MAINTAIN LOW CORROSION POTENTIAL IN HIGH-TEMPERATURE WATER

Abstract: A method for mitigating general corrosion and crack initiation and growth on the surface of metal components in a water-cooled nuclear reactor. A compound containing a non-noble metal such as zirconium or titanium is injected into the water of the reactor in the form of a solution or suspension. This compound decomposes under reactor thermal conditions to release ions/atoms of the non-noble metal which incorporate in or deposit on the surfaces of the components, including the interior surfaces of any cracks formed therein. The preferred compounds are zirconium compounds such as zirconium acetylacetonate, zirconium nitrate and zirconyl nitrate. Zirconium deposited in or on an oxided surface of a metal component will reduce the electrochemical potential at the surface to a level below the critical potential to protect against intergranular stress corrosion cracking.

Description: MODIFICATION OF OXIDE FILM ELECTRICAL CONDUCTIVITY TO MAINTAIN LOW CORROSION POTENTIAL IN HIGH-TEMPERATURE WATER Related Patent Applications This application is a continuation-in-part applica¬ tion of U.S. Patent Application No. 08/209,175 filed on March 10, 1994, which in turn is a continuation-in-part application of both U.S. Patent Application No. 08/143,513 and 08/143,514 filed on October 29, 1993. 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 β C or greater, steam, or the condensate thereof. 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 genera¬ tion, research and propulsion. A reactor pressure ves¬ sel 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 β C for a boiling water reactor (B R) , and about 15 MPa and 320°C for a pressurized water reactor (P R) . The materials used in both B Rs and P Rs 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 elec ¬ trons from anodic to cathodic areas on metallic sur- faces. The ECP is a measure of the thermodynamic ten¬ dency for corrosion phenomena to occur, and is a funda¬ mental parameter in determining rates of, e.g., SCC, corrosion fatigue, corrosion film thickening, and gen¬ eral corrosion. In a BWR, the radiolysis of the primary water cool¬ ant in the reactor core causes the net decomposition of a small fraction of the water to the chemical products H 2 , H 2 0 2 , 0 2 and oxidizing and reducing radicals. For steady-state operating conditions, equilibrium concen- trations of 0 2 , H 2 0 2 , and H 2 are established in both the water which is recirculated and the steam going to the turbine. This concentration of 0 2 , H 2 0 2 , and H 2 is oxi¬ dizing and results in conditions that can promote inter- granular stress corrosion cracking (IGSCC) of suscepti- ble materials of construction. One method employed to mitigate IGSCC of susceptible material is the applica¬ tion 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 concen¬ tration 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 oxidiz¬ ing species in the water, such as dissolved oxygen, and as a result lowers the ECP of metals in the water. How¬ ever, 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 poten¬ tial 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 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 poten¬ tial, whereas water with little or no oxidizing species present 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 feedwater. 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. 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. It has been shown that IGSCC of Type 304 stainless steel (containing 18-20% Cr, 8-10.5% Ni and 2% Mn) 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 shield¬ ing 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 stainless steel surface with palladium or other noble metal. The presence of pal¬ ladium on the stainless steel surface reduces the hydro- gen demand to reach the required IGSCC critical poten¬ tial of -230 mV(SHE) . 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 val- ues below the protection potential, i.e., -230 mV'SHE) . The manner in which this potential is achieved is imma- terial, e.g., by alloying, doping or by any other meth¬ od. 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. It was shown in later work that a thickness of -200 to 300 A 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. U.S. Patent Appln. Ser. No. 08/143,513 discloses an innovative method of in-situ application of a noble metal onto stainless steel or other metal surfaces by injecting a thermally decomposable noble metal compound into the high-temperature water that is in contact with the metal surface. That method dopes the oxide film with noble metal and provides sufficient catalytic activity for H 2 and 0 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 2 /0 2 molar ratios > 2 in the environment. ςuτmτιary of the Invention The present invention is an alternative doping method for achieving the same objective of low ECPs which result in slow or no crack growth in stainless steel and other metals. This is accomplished by doping or coating IGSCC-susceptible metal surfaces with a non- noble metal such as zirconium. The present invention polarizes the metal corrosion potential in the negative direction without the addition of hydrogen. The invention is based on reducing the electrical conductivity of the protective oxide which forms nat ¬ urally on the surfaces of components made of stainless steels and other metals in a light water nuclear reac- tor. The passive oxide films on structural materials comprise iron, nickel and chromium oxides, which are semiconducting in high-temperature water. Doping levels required to modify semiconducting behavior are typically low. This reduction in electrical conductivity forces the corrosion potential to be formed on the (inner) metal surface, not on the (outer) oxide—water interface. The poor access of oxidants in the bulk solution to the metal surface causes the corrosion potential to drop to a lower thermodynamic limit associated with deaerated water of a specific pH. Such modification of the oxide properties provides a very high degree of resistance to SCC even in the absence of any hydrogen addition. The passive oxide films on the surfaces of struc¬ tural materials can be doped or coated with zirconium or other non-noble metal using either in situ or ex situ techniques. In accordance with both techniques, the structural material is immersed in a solution or suspen¬ sion of a compound containing a non-noble metal. The non-noble metal must have the property of increasing the corrosion resistance of the stainless steel or other metal surface when incorporated therein or deposited thereon. The selected compound must have the property that it decomposes under reactor thermal conditions to release ions/atoms of the selected non-noble metal which incorporate in or deposit on the oxide film formed on the stainless steel or other metal surfaces. The pre¬ ferred compounds in accordance with the invention are those containing zirconium, e.g., the organometallic compounds zirconium acetylacetonate and inorganic compounds zirconium nitrate and .^irconyl nitrate. For the in situ technique, the concentration of zirconium in the reactor water is preferably in a range up to 100 ppb. Upon injection, the zirconium compound decomposes and deposits zirconium on the surfaces of components made of stainless steel or other metal which are immersed in the water. For example, in the case of stainless steel, the zirconium is incorporated into the oxide film on the stainless steel surface via a thermal decomposition process of the zirconium compound wherein zirconium ions/atoms apparently replace iron, nickel and/or chromium atoms in the oxide film, resulting in a zirconium-doped oxide film. The oxide film is believed to include mixed nickel, iron and chromium oxides. Alternatively, zirconium may be deposited within or on the surface of the oxide film in the form of a finely divided metal. During deposition, zirconium will be deposited inside any existing cracks on the stainless steel surfaces. The zirconium deposits around the crack mouth region and into the interior of the crack. This doping technique reduces the ECP of the stainless steel surfaces, particularly the interior surfaces of any crack, to below the critical threshold ECP. Thus, this approach should be effective against crack initiation and crack growth in stainless steel and other metals even in the absence of hydrogen in the high-temperature water environment. Furthermore, rapid depletion of zirconium from the crack interior should not occur because of the low fluid flow experienced inside the crack even if the crack mouth is in a high fluid flow regime. This promises to be a significant advantage because crack mitigation would still be achievable even if bulk surface zirconium removal occurs under high fluid flow conditions. Brief Description of the Drawings FIG. 1 is a schematic showing a partially cutaway perspective view of a conventional BWR. FIG. 2 is a plot of polarization curves for Zircaloy-2 and Zircaloy-4, illustrating a low corrosion potential of -800 mV(SHE) in 8 ppm NaN0 3 in the absence of any added hydrogen. FIG. 3 is an Auger electron spectroscopy depth profile of the surface of Type 304 stainless steel after exposure to a I M ZrO(N0 3 ) 2 , solution at 60°C for 10 days, showing that zirconium has been incorporated into the oxide film. FIG. 4 is a graph showing corrosion potential measurements for electrodes in 288°C water containing various amounts of oxygen, the electrodes being respectively made from Type 304 stainless steel (•) , Type 304 stainless steel doped with zirconium by immersion in a zirconyl nitrate solution for 10 days (π) and 20 days ( * ), and pure zirconium (o) . FIG. 5 is a graph showing corrosion potential measurements for electrodes in 550°F water containing various amounts of oxygen, the electrodes being made from Type 304 stainless steel doped with zirconium by immersion in a zirconium acetylacetonate solution for 48 hr. Detailed Description of the Preferred Embodiments The fluid flow in a boiling water reactor will be generally described with reference to FIG. 1. Feedwater is admitted into a reactor pressure vessel (RPV) 10 via a feedwater inlet 12 and a feedwater sparger 14, which is a ring-shaped pipe having suitable apertures for circumferentially distributing the feedwater inside the RPV. A core spray inlet 11 supplies water to a core spray sparger 15 via core spray line 13. The feedwater from feedwater sparger 14 flows downwardly through the downcomer annulus 16, which is an annular region between RPV 10 and core shroud 18. Core shroud 18 is a stain¬ less steel cylinder which surrounds the core 20 compris¬ ing numerous fuel assemblies 22 (only two 2 2 arrays of which are depicted in FIG. 1) . Each fuel assembly is supported at the top by top guide 19 and at the bottom by core plate 21. Water flowing through down- comer annulus 16 then flows to the core lower plenum 24. The water subsequently enters the fuel assemblies 22 disposed within core 20, wherein a boiling boundary layer (not shown) is established. A mixture of water and steam enters core upper plenum 26 under shroud head 28. Core upper plenum 26 provides standoff between the steam—water mixture exiting core 20 and entering verti- cal standpipes 30, which are disposed atop shroud head 28 and in fluid communication with core upper plenum 26. The steam-water mixture flows through standpipes 30 and enters steam separators 32, which are of the axial-flow centrifugal type. The separated liquid water then mixes with feedwater in the mixing plenum 33, which mixture then returns to the core via the downcomer annulus. The steam passes through steam dryers 34 and enters steam dome 36. The steam is withdrawn from the RPV via steam outlet 38. The BWR also includes a coolant recirculation sys¬ tem which provides the forced convection flow through the core necessary to attain the required power density. A portion of the water is sucked from the lower end of the downcomer annulus 16 via recirculation water outlet 43 and forced by a centrifugal recirculation pump (not shown) into jet pump assemblies 42 (only one of which is shown) via recirculation water inlets 45. The BWR has two recirculation pumps, each of which provides the driving flow for a plurality of jet pump assemblies. The pressurized driving water is supplied to each jet pump nozzle 44 via an inlet riser 47, an elbow 48 and an inlet mixer 46 in flow sequence. A typical BWR has 16 to 24 inlet mixers. The present invention is a technique to coat or dope stainless steel and other metal surfaces (including the interiors of cracks formed therein) inside a BWR with zirconium, titanium or other non-noble metal. In accordance with an in situ technique, this is accomp¬ lished by injecting an inorganic or organometallic compound containing zirconium, titanium or other non- noble metal into the high-temperature water of the BWR during shutdown or during operation. The invention will be disclosed with specific reference to doping of stainless steel surfaces with zirconium. However, it should be understood that other non-noble metals (such as niobium, yttrium, tungsten, vanadium, titanium, etc.) can be used in place of zirconium; and that the surfaces of components made of alloys other than stainless steel (e.g., nickel-based alloys, carbon steels, low alloy steels, etc.) can be doped to achieve reduced ECPs using the method of the present invention. Preferably the zirconium compound is injected at a point upstream of the feedwater inlet 12 (see FIG. 1) . The high temperatures as well as the gamma and neutron radiation in the reactor core act to decompose the compound, thereby freeing Zr ions/atoms for deposition on the oxide film which coats oxidized stainless steel surfaces in a BWR. One group of Zr-containing compounds which can be used for this purpose are the zirconium compounds containing nitrate groups, such as zirconyl nitrate [ZrO(N0 3 ) 2 ] and zirconium nitrate [Zr(N0 3 ) 4 ]. Another Zr-containing compound which can be used is zirconium acetylacetonate [ZrAc4]. It is known that palladium doping combined with hydrogen addition is effective in mitigating IGSCC cracking. The action of palladium doping is to cause very efficient recombination of added H 2 with 0 2 present in the system such as in an operating BWR, so that the local 0 2 levels are considerably reduced. The metal surface (e.g.. Type 304 stainless steel) in effect sees much less 0 2 , even though the bulk fluid may have a much higher 0 2 , content. The lowering of surface 0 2 (i.e., at the interface) is sufficient to bring about the necessary ECP change. The amount of H 2 required to achieve the protection potential, even if the metal surface were doped with palladium, depends to a large extent on the specific nature of the plant. As an example, in the case of a high- power-density plant, such as a BWR 4, the required hydrogen may be relatively small so that the main steam line radiation levels may still remain at the background level. However, for a low-power-density plant, such as a BWR 3, where recombination of the H 2 and 0 2 in the downcomer region is not as efficient, more H 2 may be required to achieve IGSCC protection. Thus, although palladium doping helps, the benefit may not be as much as it would otherwise be if it were a high-power-density plant. Thus, in such situations, even with palladium doping, the required H 2 levels may be sufficiently high to bring the main steam line radiation levels above the background level. On the other hand, if an element existed that could lower the ECP without the addition of hydrogen, then that would be a great benefit, because the question of increasing the main steam line radiation level does not arise. It is in this context that zirconium doping was viewed as a possible alternative to palladium doping, particularly based on observations of the ability of Z-Nb alloy to lower the ECP of a Type 304 stainless steel CERT specimen. ECP and 0 2 test data at 547°F for a Type 304 stainless steel CERT specimen held in place in a clevis using oxidized Z-Nb pins, a Type 304 stainless steel electrode tip and a Type 304 stainless steel CERT specimen held in a clevis using Zr0 2 (MgO) ceramic pins are com ¬ pared in Table I. All stainless steel specimens had been pre-oxidized before the test. TABLE I o 2 ECP Specimen (ppb) (mV,SHE) 304 SS (CERT) with oxidized Zr—Nb pins 225 -196 304 SS electrode tip 225 +60 304 SS (CERT) with ZrO.(MgO) pins 235 +90 A constant extension rate tensile (CERT) test was performed at 547 β F with a Type 304 stainless steel specimen. The specimen was held in the clevis of a standard CERT autoclave using oxidized Zr—Nb pins. During the test it was discovered that the ECP of the stainless steel specimen was far more negative (-196 mV/SHE) than expected at the oxygen level (225 ppb 0 2 ) used in the study. A preoxidized Type 304 stainless steel electrode tip that was in the same autoclave showed a potential of +60 mV(SHE) , which was antici¬ pated in the high-oxygen environment used. Based on this result, it was concluded that the oxide film formed on the Zr—Nb pin cracked under the load during the CERT test, which exposed the bare zirconium/niobium metal that contacted the Type 304 stainless steel specimen. This caused a mixed potential to be established at the stainless steel specimen dominated by the negative potential of the Zr—Nb alloy material. Thus, the Type 304 stainless steel CERT specimen showed a negative potential of -196 mV(SHE) instead of showing a positive potential at 225 ppb 0 2 . This observation is consistent with the finding that Zircaloy-2 and Zircaloy-4 also show very negative potentials of -820 mV(SHE) at 550°F in 8 ppm NaN0 3 in the presence of 248 ppb 0 2 , as shown in FIG. 2. FIG. 2 is a plot of polarization curves for Zir- caloy-2 and Zircaloy-4 illustrating a low corrosion potential of —800 mV(SHE) in 8 ppm NaN0 3 at 289°C in the absence of any added 1^, i.e., having only an oxygen level of 248 ppm 0 2 . FIG. 2 indicates that if the metal surface incorporates an insulating film such as that formed by zirconium or one of its alloys, then it is possible to achieve very low ECPs without adding any H 2 . Thus, in principle, because of the low ECPs, it should be possible to mitigate IGSCC of susceptible alloys such as Type 304 stainless steel, nickel alloys and other steels by doping the alloy surface with a metallic element such as zirconium, so that the surface behaves more like a zirconium surface, resulting in a lower ECP. Zirconium doping of stainless steel surfaces could be achieved using zirconium compounds such as zirconium acetylacetonate, zirconyl nitrate [Zr0(N0 3 ) 2 ] and zirconium nitrate [Zr(N0 3 ) 4 ]. Other dopants that can potentially be used for generating similar insulating or semiconducting surfaces include niobium, yttrium, tungsten, vanadium, titanium, molybdenum, chromium and nickel. Thus, when a metal such as stainless steel is ohm- ically shunted to a zirconium-containing alloy, the po- tential of the stainless steel polarizes in the negative direction as observed in the above study. The exact potential to which stainless steel polarizes depends upon the area ratio of the zirconium/stainless steel and the electrochemical activity of the particular zirconium alloy itself. The fact that the Zr—Nb pins in fact are polarizing the Type 304 stainless steel CERT specimen potential in the negative direction was proven by re¬ placing the Zr— b pins with ceramic (Zr0 2 (MgO)) pins, which eliminated the mixed potential effect, and result- ed in a positive potential (+90 mV(SHE)) for the stain¬ less steel CERT specimen, as expected. An important benefit of this observation is the ability to achieve required IGSCC protection potentials for Type 304 stainless steel, either by shunting the Type 304 stainless steel to zirconium or one of its alloys (electrochemically short range) , or by doping the stainless steel surface with zirconium or one of its alloys. Zirconium doping of stainless steel (or other metal) components of a BWR by injecting a zirconium compound into the high-temperature water would make it possible to polarize the stainless steel potential in the negative direction without using hydrogen. The benefits of this achievement would be numerous. First, the main steam radiation dose rates should remain at the background level because no hydrogen will be used. Second, zirconium and its alloys are compatible with fuel cladding material and hence fuel removal may not be required during zirconium doping. Lastly, the cost of zirconium is much less than the cost of palladium. In accordance with the invention, zirconium doping can be performed in situ either during shutdown (when the water temperature inside the reactor is about 40- 60 β C) or during operation (when the water temperature inside the reactor is about 288°C). As a result of injecting the solution of zirconium compound into the feedwater, all structural surfaces exposed to the recir¬ culating water carrying the injected zirconium will be doped. Alternatively, structural reactor components can be treated ex situ before installation in the reactor. An experiment was performed to test the effect on corrosion potential of exposing Type 304 stainless steel to a Zr0(N0 3 ) 2 solution. Test specimens of Type 304 stainless steel (W" diam. x 2" long) were pre-oxidized in 288 β C water containing 200 ppb oxygen for 1 week and then immersed in an ultrasonic bath containing a 1 mtf ZrO(N0 3 ) 2 solution, open to air, at 60 β C for 10 and 20 days respectively. FIG. 3 shows an Auger electron spectroscopy depth profile of the surface of Type 304 stainless steel after exposure to a 1 m ZrO(N0 3 ) 2/ solution at 60°C for 10 days. The data in FIG. 3 confirm that zirconium has been incorporated into the oxide film as a result of the treatment in accordance with the invention. Zirconium was incorporated into the oxide film to a depth of 300 A (1 minute sputtering time ~ 100 A) . The corrosion potentials of the Z-doped Type 304 stainless steel test specimens, as well as undoped and pure zirconium test specimens, were measured in 288°C water containing various amounts of oxygen. All test specimens were exposed to 288°C pure water for 2 days and the corrosion potentials were sequentially measured with incremental additions of oxygen. The experimental data is shown in FIG. 4 for Type 304 stainless steel (0), Type 304 stainless steel treated with a 1 mM ZrO(N0 3 ) 2 solution for 10 days ( □ ) and 20 days (±), and pure zirconium (0) electrodes. It was observed that the Z-doped Type 304 stainless steel test specimens showed lower corrosion potentials than the undoped specimens at the same oxygen level. This difference in the corrosion potentials of the Z-doped and undoped stainless steel electrodes is attributable to the change in the electrical conductivity of the oxide film caused by doping of zirconium into the oxide. By contrast, the corrosion potential of pure zirconium was about -650 V(SHE), even at high oxygen levels. As seen in FIG. 4, the corrosion potential of Z-doped Type 304 stainless steel is further reduced as the duration of the doping treatment is increased from 10 days to 20 days. Another experiment was performed to test the effect on corrosion potential of exposing Type 304 stainless steel to a zirconium acetylacetonate solution. The Type 304 stainless steel specimen was treated with 100 ppb Zr added as zirconium acetylacetonate in high-tempera¬ ture water (550 β F) over a period of 48 hr. An exemplary zirconium acetylacetonate injection solution was prepared by dissolving 52.6 mg of zirconium acetylacetonate powder in 40 ml of ethanol. The ethanol solution was then diluted with water. After dilution, 10 ml of ethanol are added to the solution. This sol¬ ution is then diluted with water to a volume of 1 liter. Obviously, the concentration range can be varied. Alternatively, a water-based suspension can be formed, without using ethanol, by mixing zirconium acetylace¬ tonate powder in water. The zirconium acetylacetonate compound, dissolved in the ethanol/water mixture, was injected into the inlet side of the main pump in the flow loop using an injection pump at a rate so that the solution entering the autoclave (at 550 β F) had a Zr concentration of -100 ppb. The results of this experiment are depicted in FIG. 5. After doping with zirconium acetylacetonate, the corrosion potential of the Type 304 stainless steel specimen was tested in high-temperature water at 550°F. The response of the Zr-doped specimen was tested at different oxygen levels. FIG. 5 shows that the ECP of the Zr-doped specimen was negative from the outset even in the presence of high oxygen levels. As the oxygen content was increased, the ECP decreased, which is the reverse of the effect which would be seen in undoped Type 304 stainless steel. This may be partly due to the consumption of oxygen by the deposited organics, causing a local reduction in oxygen around the stainless steel surface. This indicates that even in the absence of any hydrogen, the presence of zirconium on and in the oxide film has a beneficial influence in providing lower ECPs. As an example of the foregoing, the ECP of an undoped Type 304 stainless steel specimen pre-oxidized in 8 ppm 0 2 for one week at 550°F drops to a value of only -39 mV(SHE) even when the H 2 /0 2 molar ratio is increased to 8.5. In contrast, the Type 304 stainless steel specimen doped with zirconium acetylacetonate shows a negative potential of -87 mV(SHE) at a dissolved oxygen concentration of 338 ppb without any hydrogen. Thus, zirconium doping of the stainless steel surface is extremely beneficial in reducing the ECP of the specimen and hence in controlling crack initiation and growth in stainless steel, since ECP is a primary factor that controls SCC of stainless steel and other alloys used in a nuclear reactor. The foregoing method has been disclosed for the purpose of illustration, variations and modifications of the disclosed method will be readily apparent to practitioners skilled in the art of mitigating stress corrosion cracking in metals and alloys. For example, the non-noble metals identified above as being useful in the invention can be used alone or in combination. Also the doping technique of the invention is not restricted to use with stainless steel surfaces, but also has application in reducing the ECP of other metals which are susceptible to IGSCC, e.g., nickel- based alloys, carbon steel and low alloy steels. An alternative application technology includes having the metal compound as pressed pellets in a basket hung inside the reactor at different locations and operating the reactor with pump heat until metal doping occurs. Another approach would be to inject the compound locally into areas that have a higher susceptibility to IGSCC. All such variations and modifications are intended to be encompassed by the claims set forth hereinafter.

Claims: Clai s 1. A method for treating a metal component to mitigate cracking in a surface of said metal component during use in a water-cooled nuclear reactor or associated equipment, comprising the steps of: forming an oxide film on said surface of said metal component; and doping said oxide film with ions/atoms of a non- noble metal selected from the group consisting of zirconium, niobium, yttrium, tungsten, vanadium, titanium, molybdenum, chromium and nickel. 2. The method as defined in claim 1, wherein said doping step is performed by immersing said oxide film in a solution of a zirconium-containing compound. 3. The method as defined in claim 2, wherein said zirconium-containing compound is zirconium acetylace¬ tonate. 4. The method as defined in claim 2, wherein said zirconium-containing compound is zirconium nitrate. 5. The method as defined in claim 2, wherein said zirconium-containing compound is zirconyl nitrate. 6. The method as defined in claim 1, wherein said doping step is performed by injecting a solution of a thermally decomposable compound containing said non- noble metal into the water of said reactor while said water is being recirculated during shutdown of said reactor. 7. The method as defined in claim l, wherein said doping step is performed by injecting a solution of a thermally decomposable compound containing said non- noble metal into the water of said reactor during normal operation of said reactor. 8. The method as defined in claim l, wherein said metal component is made of stainless steel. 9. The method as defined in claim l, wherein said metal component is made of an alloy selected from the group consisting of stainless steel, nickel-based alloy, carbon steel and low alloy steel. 10. The method as defined in claim 1, wherein said oxide film is doped with said non-noble metal to a doping level sufficient to decrease the electrochemical potential on said metal component surface or inside a crack formed in said metal component surface to a level below the critical potential required to protect against intergranular stress corrosion cracking. 11. A method for mitigating general corrosion and initiation or propagation of a crack in a metal compo¬ nent having an oxided surface, comprising the step of: immersing said metal component in a solution or suspension of a compound containing a non-noble metal, wherein said non-noble metal compound has the property of being thermally decomposed to release non- noble metal ions/atoms which incorporate in or deposit on said oxided surface, and said non-noble metal is selected from the group consisting of zirconium, nio¬ bium, yttrium, tungsten, vanadium, titanium, molybdenum, chromium and nickel. 12. The method as defined in claim 11, wherein said non-noble metal is zirconium. 13. The method as defined in claim 11, wherein said non-noble metal-containing compound is zirconium acetylacetonate. 14. The method as defined in claim 11, wherein said non-noble metal-containing compound is zirconium nitrate. 15. The method as defined in claim 11, wherein said non-noble metal-containing compound is zirconyl nitrate. 16. The method as defined in claim 11, wherein said immersing step is performed by injecting a solution of a thermally decomposable compound containing said non-noble metal into the water of said reactor while said water is being recirculated during shutdown of said reactor. 17. The method as defined in claim 11, wherein said immersing step is performed by injecting a solution of a thermally decomposable compound containing said non-noble metal into the water of said reactor during normal operation of said reactor. 18. The method as defined in claim 11, wherein said metal component is made of stainless steel. 19. The method as defined in claim 11, wherein said metal component is made of an alloy selected from the group consisting of stainless steel, nickel-based alloy, carbon steel and low alloy steel. 20. The method as defined in claim 11, wherein said oxided surface is doped with said non-noble metal to a doping level sufficient to decrease the electro¬ chemical potential on said metal component surface or inside a crack formed in said metal component surface to a level below the critical potential required to protect against intergranular stress corrosion cracking when said metal component is installed in a water-cooled nuclear reactor.

Inventors: ANDRESEN PETER LOUIS, DIAZ THOMAS POMPILIO, HETTIARACHCHI SAMSON, KIM YOUNG JIN, ANDRESEN, PETER, LOUIS, DIAZ, THOMAS, POMPILIO, HETTIARACHCHI, SAMSON, KIM, YOUNG, JIN
Applicants: GEN ELECTRIC, GENERAL ELECTRIC COMPANY
Application year: 1995
Main classification: G21C17
IPC codes: C23F15, G21C17, G21C19, G21D1, G21D3