PWR Primary Water Research Articles

Effects of dissolved hydrogen on fuel crud deposition and subcooled nucleate boiling in PWR primary water at 328 °C

This work aimed to evaluate the effects of dissolved hydrogen (DH) on the crud deposition and subcooled nucleate boiling (SNB) behavior of fuel cladding tubes in a simulated primary coolant of a pressurized water reactor at 328 °C and 13 MPa. The SNB characteristic occurring on the fuel cladding tubes was periodically observed using an acoustic emission method during the deposition tests. There was little difference in the energy distribution and hit number of the boiling acoustic signals at three difference DH concentrations ranging from 0 to 70 cm3/kg H2O. The difference in the amount of crud particles on the cladding tube at each DH condition was also minor. The effect of DH on crud deposition is discussed from the viewpoint of the SNB behavior, the solubility of Ni and Fe, and the change in the crud morphology.

Replicating PWR Primary Water Conditions in Low Pressure H2-Steam Environment to Study Alloy 600 Oxidation Processes

The effects of thermodynamic parameters associated with low pressure H2-steam on the oxidation behavior of Nickel have been investigated to simulate PWR primary water conditions. An Y2O3-ZrO2 solid state reference electrode operated between 372 and 480°C, and exposed to variable hydrogen partial pressures, was used to obtain in-situ potential measurements to identify the Ni/NiO transition. Microstructural analyses on Alloy 600 coupons exposed to H2-steam under different oxidizing conditions showed a marked dependence of the intergranular oxidation processes, as expected for a PWR environment. Overall, the H2-steam simulated an environment that can be considered to be representative of PWR primary water.

An insight into PWR primary water SCC mechanisms by comparing surface and crack oxidation

Oxidation and stress corrosion cracking (SCC) of 316L stainless steel were studied in simulated pressurized water reactor primary water. Surface, crack flank, and crack tip oxides were analyzed and compared by high-resolution characterization, including oxidation state mapping. All oxides were found to have a triplex structure, although of different dimensions and composition, revealing the effects of local water chemistry and applied stress. The higher oxidation rate at the crack tip could be explained due to the existence of a higher dislocation density, higher level of stress and cation unavailability from the environment. The implications to SCC mechanisms are discussed.

Irradiation assisted stress corrosion cracking of commercial and advanced alloys for light water reactor core internals

Thirteen alloys including high- and low-strength nickel-base alloys, austenitic stainless steels, and ferritic alloys were irradiated using 2 MeV protons to a damage level of 2.5 dpa at 360 °C and assessed for their susceptibility to irradiation assisted stress corrosion cracking (IASCC) in both BWR normal water chemistry (NWC) and PWR primary water. Cracking susceptibility was highest for high strength nickel-base alloys, followed by the low strength nickel-base alloys and then the low strength iron-base alloys. Cracking in the nickel-based alloys was worst in normal water chemistry, which was reversed for the iron-based alloys. In general, cracking correlated with the degree of microstructure changes, though no single feature could be linked to cracking. IGSCC occurred in both the unirradiated and irradiated conditions in high strength nickel-base alloys with susceptibility being considerably higher following irradiation. In all cases, slip was planar, and the degree of slip localization correlated with the probability of IG crack initiation. Low strength nickel-base alloys showed the same dependence on environment as high strength alloys but were considerably less susceptible to IASCC initiation. Among the low strength iron-base alloys, alloy 800 was most susceptible to IASCC initiation in both BWR NWC and PWR primary water, which also correlated with grain boundary chromium depletion and silicon segregation. Across all alloys, cracking correlated with both the degree of localized deformation and the hardness in the irradiated condition. The agreement is expected as increased hardening also correlates with localized deformation, which is likely a necessary, though insufficient condition for cracking.

Effect of electropolishing on general corrosion of Alloy 690TT tubes in simulated primary coolant of pressurized water reactors

The purpose of this study is to investigate the effect of electropolishing (EP) time on the general corrosion of Alloy 690TT steam generator tubes in a simulated PWR primary water of at 330 °C. The surface roughness and nanohardness of the tubes gradually decreased, and the electron work function and water contact angle gradually increased with increasing the EP time. However, the effect of EP time on the surface properties mainly changed after 8 min. The corrosion rate was reduced through EP treatment by approximately 60% compared to that of the commercial as-received tube.

Crack growth and cracking behavior of Alloy 600/182 and Alloy 690/152 welds in simulated PWR primary water

The crack growth responses of as-received and as-welded Alloy 600/182 and Alloy 690/152 welds to constant loading were measured by a direct current potential drop method using compact tension specimens in primary water at 325 °C simulating the normal operating conditions of a nuclear power plant. The as-received Alloy 600 showed crack growth rates (CGRs) between 9.6 × 10 −9 mm/s and 3.8 × 10 −8 mm/s, and the as-welded Alloy 182 had CGRs between 7.9 × 10 −8 mm/s and 7.5 × 10 −7 mm/s within the range of the applied loadings. These results indicate that Alloys 600 and 182 are susceptible to cracking. The average CGR of the as-welded Alloy 152 was found to be 2.8 × 10 −9 mm/s. Therefore, Alloy 152 was proven to be highly resistant to cracking. The as-received Alloy 690 showed no crack growth even with an inhomogeneous banded microstructure. The cracking mode of Alloys 600 and 182 was an intergranular cracking; however, Alloy 152 was revealed to have a mixed (intergranular + transgranular) cracking mode. It appears that the Cr concentration and the microstructural features significantly affect the cracking resistance and the cracking behavior of Ni-base alloys in PWR primary water.

Stress Corrosion Behaviors of 316LN Stainless Steel in a Simulated PWR Primary Water Environment.

The effect of the strain rate, experimental temperature, Zn content in the test solution, and prefilming time on the mechanical properties was investigated by a tensile test with a slow strain rate, at a chemical solution of 2.2 ppm Li and 1200 ppm B in a static autoclave with 8.2 MPa. The experimental parameters clearly affected the tensile properties. The surface morphology, fractograph, and cross-sectional microstructure were analyzed by scanning electron microscopy and transmission electron microscopy. The δ (elongation) and UTS (ultimate tensile strength) of the samples tested in chemical solution were obviously lower than those of the samples tested under a nitrogen atmosphere. However, in general, all samples showed a ductile fracture characteristic and an excellent tensile property in all experimental conditions. The δ and UTS were first increased with increasing Zn content, and then decreased at both conditions of 9.26 × 10−7/s and 4.63 × 10−7/s strain rates. The difference values of tensile properties at different strain rates showed fluctuations with increasing Zn content. The δ increased with both increasing experimental temperature and prefilming time. The UTS first decreased with increasing prefilming time and then increased. The Iscc (stress corrosion cracking susceptibility) decreased with an increasing strain rate, experiment temperature, and prefilming time. Many particles with polyhedrons were formed on the sample surfaces, which was attributed to corrosion in a periodical location at the sample surface. The average length of the particles decreased with increasing Zn content, but increased with both increasing experimental temperatures and prefilming time. The corresponding mechanism is also discussed in this work.

Effect of aluminum oxide coated on fuel cladding surface on crud deposition in simulated PWR primary water

Deposits on fuel assemblies (crud) have been the main cause of local power shift, increased fuel cladding temperature, accelerated fuel cladding corrosion, and radiation buildup in nuclear power plants. We report the experimental results on aluminum oxide (Al2O3) as a fuel cladding coating for crud mitigation. Crud deposition tests were performed under a sub-cooled nucleate flow boiling condition in simulated primary water of pressurized water reactors at 328 °C. The amount of deposits on the Al2O3-coated cladding tube decreased by 23% compared with that on the uncoated cladding tube. The difference of zeta potentials between magnetite particle and cladding surface increased from 7.5 mV to 16.1 mV, when the cladding surface was coated with Al2O3. Therefore, the reduction of crud deposition can be attributed to the increased repulsive force between the magnetite particle and Al2O3-coated cladding surface. The effect of the Al2O3 layer is discussed from the view point of zeta potential, wettability, and thermodynamic stability.

EBSD Characterization of IGSCC Crack Initiated on Alloy 600 in Simulated PWR Primary Water Environment

In the present work, we crystallographically examined intergranular cracks initiated on Alloy 600 in a simulated PWR environment by using EBSD. The cracks initiated on a flat tensile specimen of the alloy in a short period of slow strain rate tests (SSRTs), that is, the initiation of cracks was accelerated due to the asperity of the specimen surface induced by mechano-chemical polishing with colloidal silica suspension prior to the SSRTs. As thick oxide films were formed on the specimen surface during the SSRTs, the sputtering of the films was carried out for the removal of the film. After the sputtering, clear crystallographic information such as inverse pole figures (IPFs) on the specimen surface could be obtained. Based on the information, the intergranular cracks were crystallographically analyzed. Furthermore, three-dimensional characterization of the cracks was also performed.

Effects of iron content in Ni[sbnd]Cr[sbnd]Fe alloys on the oxide films formed in an oxygenated simulated PWR water environment

The iron content in NiCrFe alloys strongly affected the properties of oxide films after 1012 h of immersion in an oxygenated simulated PWR primary water environment at 310 °C. Increasing the iron content in the alloy increased the porosity and thickness of the outer layer and the inner layer. Increasing the iron content in the alloy increased the amount of spinel needle-like oxides and decreased the amount of NiO particles in the outer layer. Cellular oxide in the inner layer was nickel-rich in the center and chromium-rich at the edge.

A mechanistic study of SCC in Alloy 600 through high-resolution characterization

High-resolution characterization was used to understand the mechanisms controlling stress corrosion cracking (SCC) in Alloy 600 exposed to simulated PWR primary water conditions. Three potential active crack tips obtained from different types of grain boundaries were studied and compared. The results suggest that the dominant mechanism controlling SCC propagation is intergranular internal oxidation. The applied stress, pre-existent residual strain, the accumulation of defects around the crack tip, the formation of a Fe-Cr-depleted zone, and a porous intergranular oxide are acknowledged as necessary precursors to SCC. Based on the results obtained in this study, a model of SCC propagation is proposed.

The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

The effect of cold rolling on grain boundary structure and stress corrosion cracking susceptibility of twins in alloy 690 in simulated PWR primary water environment

The effects of cold rolling on the structure and stress corrosion cracking (SCC) susceptibility of twin boundaries in alloy 690 were studied. Most of the twin boundaries were transformed into random high angle boundaries (RHABs) following cold rolling. The transformed twin boundaries (TTBs) showed increased susceptibility to carbide precipitation during subsequent aging at 475°C. The SCC initiation test in 360°C hydrogenated water indicated that the prior RHABs exhibited increased SCC susceptibility after cold rolling. Moreover, TTBs become susceptible to SCC due to the promoted outward diffusion of Cr and can enhance the connectivity of susceptible grain boundaries.

Effects of iron content in Ni-Cr-xFe alloys and immersion time on the oxide films formed in a simulated PWR water environment

The iron content in Ni-Cr-xFe (x = 0–9 at.%) alloys strongly affected the properties of oxide films after 978 h of immersion in the simulated PWR primary water environment at 310 °C. Increasing the iron content in the alloys increased the amount of iron-bearing polyhedral spinel oxide particles in the outer oxide layer and increased the local oxidation penetrations into the alloy matrix from the chromium-rich inner oxide layer. The effects of iron content in the alloys on the oxide film properties after 500 h of immersion were less significant than those after 978 h. Iron content increased, and chromium content decreased, in the outer oxide layer with increasing iron content in the alloys. Increasing the immersion time facilitated the formation of the local oxidation penetrations along the matrix/film interface and the nickel-bearing spinel oxides in the outer oxide layer.

Corrosion fatigue and microstructural characterisation of Type 316 austenitic stainless steels tested in PWR primary water

Fatigue crack growth experiments have been performed on high and low sulphur Type 316 austenitic stainless steel specimens in simulated Pressurised Water Reactor primary coolant environments and evaluated via microstructural characterisation techniques to further the understanding of the mechanistic behaviour. At relatively high loading frequencies, the enhanced crack growth for both specimens appeared to be crystallographic and associated with slip localization. However, when the loading frequency was decreased, the crack growth rates for the low S specimen increased, whereas the high S specimen exhibited retarded crack growth and the crack path was no longer crystallographic.

Effect of irradiation on corrosion of 304 nuclear grade stainless steel in simulated PWR primary water

Effect of irradiation on corrosion of 304 nuclear grade stainless steel (304NG SS) in simulated primary water of pressurized water reactor was investigated. The investigation was conducted by comparing the microstructure of the oxide scale formed on the steel irradiated to different doses by proton. The results revealed that increasing irradiation dose promoted both the corrosion and intergranular corrosion of 304NG SS. This is attributed to the irradiation induced defects as well as the Cr depletion at the grain boundary in the steel.

Boric Acid Radiolysis in Primary Coolant Water of PWR at Temperature of 250oC

BORIC ACID RADIOLYSIS IN PRIMARY COOLANT WATER OF PWR AT TEMPERATURE OF 250oC. The existence of oxygen in the primary coolant system of PWR could lead to corrosion, hence it is very important to suppress the oxygen concentration in the system. Therefore, study of the effect of boric acid addition into the primary coolant water system of PWR to suppress oxygen concentration resulted from gamma-ray radiation is essential to be performed. The aim of this research is to understand reaction mechanism at temperature of 2500C and the effect of boric acid adding toward oxygen concentration in the PWR primary coolant water. Methodology used is simulation using Facsimile software. Input for the software namely radiolysis reaction mechanism for pure water, G value from radiolysis product, dose rate of 1 and 104 Gy/s, aeration and deaeration system, and specific reaction of boric acid with hydroxyl radical and hydrated electron at temperature 250C and 3000C. The output are in the form of irradiation time vs oxygen concentration time series. The results show that the oxygen production increase significantly with the irradiation time and reach the saturated concentration at 107s. Based on the plot of oxygen’s concentration at 107s vs boric acid, several results are as following: oxygen concentration significantly suppressed by boric acid addition and gives the exponential decreasement, the higher dose rate gives the higher concentration of oxygen, the aeration system gives no effect on suppressing oxygen concentration at boric acid addition up to 0.1M.

Intergranular oxidation of 316L stainless steel in the PWR primary water environment

After exposing AISI 316L stainless steel to the primary water environment across a range of temperatures (290°C–360°C) for 1000h, the oxidation growth of the Cr-rich inner layer was measured. Samples exposed at 290°C proved to have significantly greater oxide penetration than samples exposed at the higher temperatures. Furthermore, the annealed samples exhibit strong dependence of oxide growth on microtexture. The results suggest that an epitaxial relationship exists between the metal and the Cr-rich oxide, and that this relationship degrades above 290°C as a result of differential thermal expansion between the oxide and the metal matrix.

Stress corrosion crack initiation of alloy 600 in PWR primary water

Stress corrosion crack (SCC) initiation of three mill-annealed alloy 600 heats in simulated pressurized water reactor primary water has been investigated using constant load tests equipped with in-situ direct current potential drop (DCPD) measurement capabilities. SCC initiation times were greatly reduced by a small amount of cold work. Shallow intergranular attack and/or cracks were found on most high-energy grain boundaries intersecting the surface with only a small fraction evolving into larger cracks and intergranular SCC growth. Crack depth profiles were measured and related to DCPD-detected initiation response. Processes controlling the SCC initiation in mill-annealed alloy 600 are discussed.

Characterization of interfacial reactions and oxide films on 316L stainless steel in various simulated PWR primary water environments

The effect of water chemistry on the electrochemical and oxidizing behaviors of 316L SS was investigated in hydrogenated, deaerated and oxygenated PWR primary water at 310 °C. Water chemistry significantly influenced the electrochemical impedance spectroscopy parameters. The highest charge-transfer resistance and oxide-film resistance occurred in oxygenated water. The highest electric double-layer capacitance and constant phase element of the oxide film were in hydrogenated water. The oxide films formed in deaerated and hydrogenated environments were similar in composition but different in morphology. An oxide film with spinel outer particles and a compact and Cr-rich inner layer was formed in both hydrogenated and deaerated water. Larger and more loosely distributed outer oxide particles were formed in deaerated water. In oxygenated water, an oxide film with hematite outer particles and a porous and Ni-rich inner layer was formed. The reaction kinetics parameters obtained by electrochemical impedance spectroscopy measurements and oxidation film properties relating to the steady or quasi-steady state conditions in the time-period of measurements could provide fundamental information for understanding stress corrosion cracking processes and controlling parameters.