Process to mitigate stress corrosion cracking of structural...

Details Process to mitigate stress corrosion cracking of structural... Process to mitigate stress corrosion cracking of structural...

2003-06-16

2004-04-20

Behrend, Harvey E. (Department: 3641)

Induced nuclear reactions: processes, systems, and elements

Reactor protection or damage prevention

Corrosion or damage prevention

C422S014000

Reexamination Certificate

active

06724854

ABSTRACT:

BACKGROUND
This disclosure generally relates to mitigating stress corrosion cracking of components exposed to high temperature water, and more particularly, reducing a corrosion potential of the components.
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 feed water back to the vessel. Operating pressures and temperatures for the reactor pressure vessel are about 7 Mpa and about 288° C. for a boiling water reactor (BWR), and about 15 Mpa and 320°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, low 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 oxidizing species, e.g., O
2
, H
2
O
2
, are present in the reactor water in concentrations of about 1 to 5 parts per billion (ppb) or greater. SCC is increased in a high radiation flux where oxidizing species, such as oxygen, hydrogen peroxide, and short-lived hydroxyl 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 H
2
, H
2
O
2
, and O
2
are established in both the water, which is recirculated, and the steam going to the turbine. This concentration of H
2
, H
2
O
2
, and O
2
is oxidizing and results in conditions that can promote intergranular stress corrosion cracking of susceptible materials of construction. One method employed to mitigate intergranular stress corrosion cracking 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 gaseous hydrogen to the reactor feed water. 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 and in the bulk water. The rate of these recombination reactions is dependent on local radiation fields, water flow rates, and other variables such as the oxide chemistry.
The injected hydrogen reduces the level of oxidizing species in the water, such as dissolved oxygen and hydrogen peroxide, 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.
FIG. 1
shows the observed (data points) and predicted (curves) crack growth rates as a function of corrosion potential for 25 millimeter (mm) compact tension (CT) specimens of furnace sensitized Type 304 stainless steel (containing 18-20% Cr, 8-10.5% Ni and 2% Mn) at a constant load of 25 Ksiin over the range of solution conductivities from 0.1 to 0.3 &mgr;S/cm. The data clearly shows the dependency of ECP on crack propagation rate. Data points at elevated corrosion potentials and crack propagation rates correspond to irradiated water chemistry conditions in test or commercial reactors. The shaded region at low corrosion potentials and crack propagation rates (labeled hydrogen water chemistry) correspond to normal water chemistry outside the core, i.e., non-irradiated. Thus, varying amounts of hydrogen have been required to reduce the levels 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 about −300 millivolts (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 and hydrogen peroxide 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 as shown.
Corrosion potentials of stainless steels and other structural materials in contact with reactor water containing oxidizing species can be reduced below the critical potential by injection of hydrogen into the feed water. For adequate feed water hydrogen addition rates, conditions necessary to inhibit intergranular stress corrosion cracking can be established in certain locations of the reactor. Different locations in the reactor systems 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 intergranular stress corrosion cracking of Type 304 stainless steel used in BWRs can be mitigated by reducing the ECP of the stainless steel to values below about −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 intergranular stress corrosion cracking 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 o

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