Daniel Cubicciotti

Potential-pH diagrams for alloy-water systems under LWR conditions

Abstract Potential-pH diagrams, which present the regions of stability of metallic species with water, are generally known as Pourbaix diagrams. They are especially useful to corrosion scientists because the diagrams indicate the areas of potential and pH in which oxides of the metal are stable and thus the regions in which an oxide corrosion layer can form on the metal, protecting it from corrosion. Potential-pH diagrams are generally available for the metallic elements, even at high temperature (up to about 600 K). But rarely is a pure metal used as a construction material. Stainless steel, which is an alloy of iron, chromium, and nickel plus some trace elements, is often used as a construction material for water systems because of its good corrosion resistance. The oxides in equilibrium with the water-Fe-Cr-Ni system are generally mixed metal oxides. In this paper, Pourbaix diagrams for this system are calculated from thermodynamic values at room temperature and at typical LWR (light water reactor) temperature. Diagrams for the ancilliary systems Fe-Cr-water, Fe-Ni-water and Cr-Ni-water are also given.

Equilibrium chemistry of nitrogen and potential-pH diagrams for the Fe-Cr-H2O system in bwr water

Abstract A thermodynamic analysis of the nitrogen-water system for BWR operating conditions assesses the chemical species responsible for steam line radiation increases. Nitrogen (N2) was the important gaseous form during hydrogen injection into BWR feedwater. The potential-pH diagram for the system Fe-Cr-H2O was derived from thermodynamic information for BWR operating temperatures. The diagram shows that a mixed oxide (FeCr2O4) is stable under the reducing conditions imposed by hydrogen water chemistry (HWC) for BWR while the simple oxide Fe2O3 is stable for the more oxidizing conditions of normal water chemistry. The presence of Cr in the oxide may account for the mitigation of intergranular stress corrosion cracking of sensitized stainless steel during HWC.

Vapor transport of fission products under nuclear accident conditions

Abstract Vapor transport of fission products out of the core of a nuclear reactor during a core dryout accident is modeled as a two-step process. The steps include: (1) transport throught the solid fuel to the surface and (2) vaporization into the gas stream flowing past the fuel. Transport through the solid is treated by the steam oxidation model. Calculation of the fractional release of the rate gases (Xe and Kr) from the TMI-2 accident indicated about 55% release, in good agreement with the observed releases which were in the range of 50 to 70%. The vaporization step is modeled by a set of vapor presure equations derieved from a free energy minimization treatment of the vaporization equilibrium under accident conditions. A treatment of vapor pressured of lanthanides, Ba and Sr is presented together with a simple set of equations for modeling them under accident conditions. Molybdenum vaporization is also treated. It is more complex and depends on the cesium partial pressure as well as other chemical conditions (such as H 2 pressure, H 2 O pressure and temperature) and requires a more complex model.

Flow-assisted corrosion of steel and the influence of Cr and Cu additions

Abstract Flow-assisted corrosion (FAC) of steel feedwater lines occurs by dissolution of the surface oxide layer on the steel. The solubility of iron in water under FAC conditions is discussed through the use of potential-pH diagrams (Pourbaix diagrams). Alloying additions of chromium and copper both decrease FAC. An assessment is presented that Cr additions decrease FAC by forming a mixed oxide with iron instead of a pure iron oxide. The solubility of iron from the mixed oxide is smaller than for pure iron oxide and leads to a smaller FAC rate. The stable form of copper under FAC conditions is not the mixed iron-copper oxide but metallic copper, which may act in the underlying steel surface to impede FAC.

Hydrogen water chemistry for BWRs

Abstract Intergranular stress corrosion cracking (IGSCC) has been responsible for more than 1,000 cases of cracking in austenitic stainless steel piping systems in boiling water reactors (BWRs). This paper presents the status of efforts in the United States to prevent IGSCC in BWRs during power operation by modifying the chemistry of the reactor water. The technical basis for this alternative water chemistry, called hydrogen water chemistry (HWC), is described and the results are presented of an ongoing in-plant program, at Commonwealth Edisons Dresden-2 plant, to verify the HWC concept and systematically assess the consequences of using it in an operating BWR. In addition, progress toward implementation of HWC at other U.S. plants is summarized.

Microbial-induced corrosion in nuclear power plant materials

The long construction times associated with nuclear plants and the large number of redundant or standby systems where water is allowed to remain stagnant for long periods of time produce conditions under which microbial-induced corrosion (MIC) can occur. Carbon and low-alloy steels, stainless steels and copper alloys are all susceptible to MIC in raw-water applications. Visual examination is particularly useful in performing preliminary assessments of MIC. If properly diagnosed, MIC can be effectively treated during plant construction, operation and temporary shutdowns.

Low temperature sensitization of type 304 stainless steel pipe weld heat affected zone

Large-diameter Type 304 stainless steel pipe weld heat-affected zone (HAZ) was investigated to determine the rate at which low temperature sensitization (LTS) can occur in weld HAZ at nuclear reactor operating temperatures and to determine the effects of LTS on the initiation and propagation of intergranular stress corrosion cracks (IGSCC). The level of sensitization was determined with the electrochemical potentiokinetic reactivation (EPR) test, and IGSCC susceptibility was determined with constant extension rate tests (CERT) and actively loaded compact tension (CT) tests. Substructural changes and carbide compositions were analyzed by electron microscopy. Weld HAZ was found to be susceptible to IGSCC in the as-welded condition for tests conducted in 8-ppm-oxygen, high-purity water at 288 °C. For low oxygen environments (i.e., 288 °C/0.2 ppm O2 or 180 °C/1.0 ppm O2), IGSCC susceptibility was detected only in weld HAZ that had been sensitized at temperatures from 385 °C to 500 °C. Lower temperature heat treatments did not produce IGSCC. The microscopy studies indicate that the lack of IGSCC susceptibility from LTS heat treatments below 385 °C is a result of the low chromium-to-iron ratio in the carbide particles formed at grain boundaries. Without chromium enrichment of carbides, no chromium depleted zone is produced to enhance IGSCC susceptibility.

Thermodynamics of vaporization of ZrI/sub 4/ from substoichiometric solid zirconium iodides

The pressures of gaseous ZrI/sub 4/ for solid zirconium iodides in the composition range ZrI/sub 2/ /sub 4/-ZrI/sub 1/ /sub 0/ were measured by a torsion-effusion method. A univariant equilibrium region was found between the solids ZrI/sub 1/ /sub 9/ and ZrI/sub 1/ /sub 3/ for which the pressure of gaseous ZrI/sub 4/ was given by log/sub 10/ P(atm) = -9308/T + 8.84. A review of the literature concerning the solid phase in the Zr-I system indicates that there are solid solution regions corresponding approximately to the tetra, tri, di, and monoiodide. 2 figures, 3 tables.

Thermodynamics of vaporization of fission products and materials under severe reactor accident conditions: Analysis of molten core/concrete chemistry

Abstract Vaporization-condensation processes can generate radioactive aerosols in the event of a core dryout and meltdown accident at a nuclear power station. The time sequence of fission produce vaporization and aerosol formation in relation to processes that can transport them out of the reactor containment is important for assessing their potential biohazard. Thermodynamics of vaporization of fission products and other materials are evaluated for the extreme environmental conditions projected by computer models if a molten core penetrates the reactor vessel and melts into the concrete base. A free energy minimization treatment was used to estimate partial pressures of gases in this many-component, multiphase system. The amounts of fission products and condensable materials vaporized were calculated for a test case involving basalt-aggregate concrete.

LWR water chemistry guidelines

Abstract This review describes the development of water chemistry guidelines to reduce corrosion damage and control radiation buildup in light water reactors. Three sets of guidelines are reviewed, covering BWR systems, PWR primary systems, and PWR secondary systems. The technical basis for the water chemistry parameters are discussed and the corrective actions to be taken if parameters exceed the recommended values are outlined.