Stress corrosion cracking of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The stress corrosion cracking (SCC) behaviour of different reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated boiling water reactor (BWR) normal water (NWC) and hydrogen water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 °C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (<0.6 mm/year) under most BWR/NWC and material conditions. Critical water chemistry, loading and material conditions, which can result in sustained and fast SCC well above the ‘BWRVIP-60 SCC disposition lines’ were identified, but many of them generally appeared atypical for current optimized BWR power operation practice or modern RPVs. Application of HWC always resulted in a significant reduction of SCC crack growth rates by more than one order of magnitude under these critical system conditions and growth rates dropped well below the ‘BWRVIP-60 SCC disposition lines’.     

Corrosion fatigue behavior of low alloy steels under simulated BWR coolant conditions

The corrosion fatigue crack growth behavior of A533 and A508 low alloy steels under simulated boiling water reactor (BWR) coolant conditions was studied. Corrosion fatigue crack growth rates of A533B3 and A508 cl. 3 steels were significantly affected by the steel sulfur content, loading frequency and dissolved oxygen content of water environments. The data points outside the bound of Eason’s model could be attributed to the low frequency, higher steel sulfur content and high dissolved oxygen in water environments. The sulfur dissolved in the water environment from the higher-sulfur steels was sufficiently concentrated to acidify the crack tip chemistry even in the hydrogen water chemistry (HWC). Therefore, nitrogenated or HWC water showed little or no beneficiary effect on the high-sulfur steels. For the steel specimens of the same sulfur level, their corrosion fatigue crack growth rates were comparable in different orientations, which could be related to the exposure of fresh sulfides to the water environment. The percentages of sulfides per unit area, by quantitative metallography, were comparable for the steel specimens of both orientations. When the steel sulfur content was decreased to a critical sulfur content 0.005 wt.%, the crack growth rates decreased remarkably.     

Effect of Hydrogen Water Chemistry on Zircaloy-2 Fuel Cladding and Structure Material Performance

The effects of hydrogen addition to the feedwater on the corrosion and hydrogen uptake performance of Zircaloy-2 fuel cladding tubes, a water rod tube and spacer materials irradiated for four cycles in a BWR were evaluated. The uniform oxide behaviors of the cladding tubes, water rod and spacer materials were not affected by hydrogen water chemistry (HWC) condition. The hydrogen uptake and pickup fractions of the water rod and spacer materials were similar to those of water rods and spacer materials under normal water chemistry (NWC) conditions. As for the fuel rods, in spite of comparably heavy crud deposition, their hydrogen uptake and pickup fractions were clearly lower than the values under NWC conditions. Overall, the results indicated that HWC had no adverse effects on fuel performance.     

Effects of hydrogen water chemistry on radiation field buildup in BWRs

A number of boiling water reactor (BWR) plants worldwide are currently operating under hydrogen water chemistry (HWC). In some reactors, when switching from normal water chemistry (NWC) to HWC, an increase in the recirculation piping dose rates has been observed. Understanding the key factors which affect the dose rate increase is the subject of our current investigation. Laboratory experiments have been conducted under control chemistry conditions to examine the rates of ⁶⁰Co deposition and the characteristic of oxide films formed on stainless steel surfaces. The activity buildup data obtained from two operating BWRs are carefully reviewed and discussed in this paper. Based on both laboratory and reactor data, a plausible mechanism of enhanced activity buildup under HWC conditions is hypothesized.     

Predicted impact of core flow rate on the corrosion mitigation effectiveness of hydrogen water chemistry for Kuosheng boiling water reactor

It is currently a common practice that a boiling water reactor (BWR) adopts hydrogen water chemistry (HWC) for mitigating corrosion in structural components in its primary coolant circuit. When the core flow rate (CFR) in a BWR is changed, the coolant residence time in the primary coolant circuit would be different. The concentrations of major redox species (i.e. hydrogen, oxygen, and hydrogen peroxide) in the coolant may accordingly vary due to different durations of radiolysis in the core and other near-core regions. A theoretical model by the name of DEMACE was used in the current study to investigate the impact of various CFRs (from 100% to 80.0%) on the effectiveness of HWC in a domestic BWR. Our analyses indicated that the HWC effectiveness at some locations could be downgraded due to a decrease in CFR. However, a lower CFR was instead beneficial to the corrosion mitigation efficiency of HWC at other locations. The impact of CFR on the HWC effectiveness could vary from location to location in a BWR and eventually from plant to plant.     

Predicted impact of power uprate on the water chemistry of Kuosheng boiling water reactor

A theoretical model was adapted to evaluate the impact of power uprate on the water chemistry of a commercial boiling water reactor (BWR) in this work. In principle, the power density of a nuclear reactor upon a power uprate would change immediately, followed by water chemistry variations due to enhanced radiolysis of water in the core and near-core regions. It is currently a common practice for commercial BWRs to adopt hydrogen water chemistry (HWC) for corrosion mitigation. The optimal feedwater hydrogen concentration may be different after a power uprate is implemented in a BWR. A computer code DEMACE was used in the current study to investigate the impact of various power uprate levels on major radiolytic species concentrations and electrochemical corrosion potential (ECP) behavior of components in the primary coolant circuit of a domestic BWR-6 type reactor operating under either normal water chemistry or HWC. Our analyses indicated that under a constant core flow rate the chemical species concentrations and the ECP did not vary monotonously with increases in reactor power level at a fixed feedwater hydrogen concentration. In particular, the upper plenum and the upper downcomer regions exhibited uniquely higher ECPs at 108% and 115% power levels than at the other evaluated power levels.     

Stress corrosion cracking of low-alloy, reactor-pressure-vessel steels in oxygenated, high-temperature water

The stress corrosion cracking (SCC) behaviour of low-alloy, reactor-pressure-vessel (RPV) steels in oxygenated, high-temperature water and its relevance to boiling water reactor (BWR) power operation, in particular its possible effect on both RPV structural integrity and safety, has been a subject of controversial discussions for many years. This paper presents the results of an experimental study on crack growth through SCC in three, nuclear-grade, steels (SA 533 B Cl.1, SA 508 Cl.2, 20 MnMoNi 5 5) under simulated, BWR water-chemistry conditions. Modern, high-temperature water loops, on-line crack-growth monitoring and fractographic analysis in the scanning electron microscope were used to quantify the cracking response of pre-cracked, fracture-mechanics specimens under a variety of mechanical and environmental conditions. Corrosion-assisted crack advance could be only initiated by active loading within the environment. If SCC crack advance at constant load was observed, initiation of crack growth had always occurred while increasing the load to the intended value for subsequent, static-load testing. During the constant load period the rate of SCC crack advance rapidly decayed and crack arrest occurred within a period of <100 h (for tests with K_I60 MPa m^1/2). Supplementary experiments with slowly increasing loading revealed that the initiation of crack growth, and the extent of further crack advance, are crucially dependent upon maintaining both a positive crack-tip strain rate and a high sulphur-anion activity in the crack-tip environment. It is concluded that there is no sustainable susceptibility to SCC crack growth under purely static loading, as long as small-scale-yielding conditions prevail at the crack-tip and the water chemistry is maintained within current BWR/NWC operational practice (EPRI water chemistry guidelines). However, sustained, fast SCC (with respect to operational time scales) cannot be excluded for faulted water-chemistry conditions (>EPRI action level 3) and/or for highly stressed specimens either loaded near to K_IJ or with a high degree of plasticity in the remaining ligament.     

Strain-induced corrosion cracking behaviour of low-alloy steels under boiling water reactor conditions

The strain-induced corrosion cracking (SICC) behaviour of different low-alloy reactor pressure vessel (RPV) and piping steels and of a RPV weld filler/weld heat-affected zone (HAZ) material was characterized under simulated boiling water reactor (BWR)/normal water chemistry (NWC) conditions by slow rising load (SRL) and very low-frequency fatigue tests with pre-cracked fracture mechanics specimens. Under highly oxidizing BWR/NWC conditions (ECP +50 mV_SHE, 0.4 ppm dissolved oxygen), the SICC crack growth rates were comparable for all materials (hardness <350 HV5) and increased (once initiated) with increasing loading rates and with increasing temperature with a possible maximum/plateau at 250 °C. A minimum K_I value of 25 MPa m^1/2 had to be exceeded to initiate SICC in SRL tests. Above this value, the SICC rates increased with increasing loading rate dK_I/dt, but were not dependent on the actual K_I values up to 60 MPa m^1/2. A maximum in SICC initiation susceptibility occurred at intermediate temperatures around 200–250 °C and at slow strain rates in all materials. In contrast to crack growth, the SICC initiation susceptibility was affected by environmental and material parameters within certain limits.     

In situ measurement of corrosion of type 316L stainless steel in 553 K pure water via the electrical resistance of a thin wire

A system for the in situ monitoring of corrosion depth via electrical resistance measurements was applied to study the corrosion rate of type 316L stainless steel at 553 K in pure water. Corrosion depth was measured using a 50 μm diameter wire probe mounted axially in the tube. Measurements were in good agreement with literature data for both the hydrogen water chemistry (HWC) condition and the normal water chemistry (NWC) condition. Oxide film analyses by scanning electron microscopy and laser Raman spectroscopy on the wire probe and the tube showed no effects from shape of the test specimens or the application of electric current. Corrosion kinetics was evaluated by fitting equations to the measurements. Data for the HWC condition could be fitted by a two-step logarithmic–parabolic law. A single-step logarithmic law fitted data for the NWC condition. Changes in corrosion rate by the water chemistry changes were readily detected with the technique. Corrosion depth change could be observed for the water chemistry change from the NWC condition to the HWC condition with electrochemical corrosion potential (ECP) of ?0.56 V vs. standard hydrogen electrode, which is lower than the ECP that the phase of iron oxide changes from α-Fe₂O₃ to Fe₃O₄.     

Modeling of SCC initiation and propagation mechanisms in BWR environments

During operation of mainly BWRs’ (Boiling Water Reactors) excursions from recommended water chemistries may provide favorite conditions for stress corrosion cracking (SCC). Maximum levels for chloride and sulfate ion contents for avoiding local corrosion are therefore given in respective water specifications. In a previously published deterministic 288 °C – corrosion model for Nickel as a main alloying element of BWR components it was demonstrated that, as a theoretically worst case, bulk water chloride levels as low as 30 ppb provide local chloride ion accumulation, dissolution of passivating nickel oxide and precipitation of nickel chlorides followed by subsequent local acidification. In an extension of the above model to SCC the following work shows that, in a first step, local anodic path corrosion with subsequent oxide breakdown, chloride salt formation and acidification at 288 °C would establish local cathodic reduction of accumulated hydrogen ions inside the crack tip fluid. In a second step, local hydrogen reduction charges and increasing local crack tip strains from increasing crack lengths at given global stresses are time stepwise calculated and related to experimentally determined crack critical cathodic hydrogen charges and fracture strains taken from small scale SSRT tensile tests pieces. As a result, at local hydrogen equilibrium potentials higher than those of nickel in the crack tip solution, hydrogen ion reduction initiates hydrogen crack propagation that is enhanced with increasing global stresses. In accordance with respective experimental literature data it is shown that decreasing chloride and increasing pH levels of the primary bulk water at 288 °C reduce the total crack propagation rates including anodic path corrosion as well as hydrogen cracking. It is also demonstrated that crack propagation rates can be significantly suppressed by hydrogen water chemistry (HWC) that leads to reduction of bulk surface corrosion potentials. As a conclusion the extended SSC-model for nickel supplies quantitative insight into the frequently controversially discussed high temperature SCC mechanisms of a basic alloying element of BWR components.     

Hydrogen water chemistry for BWRs: a status report on the EPRI development program

Many boiling water reactors (BWRs) have experienced extensive intergranular stress corrosion cracking (IGSCC) in their austenitic stainless steel reactor coolant system piping, resulting in serious adverse impacts on plant capacity factors, O&M costs, and personnel radiation exposures. A major research program to provide remedies for BWR pipe cracking was co-funded by EPRI, GE, and the BWR Owners Group for IGSCC Research between 1979 and 1988. Results from this program show that the likelihood of IGSCC depends on reactor water chemistry (particularly on the concentrations of ionic impurities and oxidizing radiolysis products) as well as on material condition and the level of tensile stress. Tests have demonstrated that the concentration of oxidizing radiolysis products in the recirculating reactor water of a BWR can be reduced substantially by injecting hydrogen into the feedwater. Recent plant data show that the use of hydrogen injection can reduce the rate of IGSCC to insignificant levels if the concentration of ionic impurities in the reactor water is kept sufficiently low. This approach to the control of BWR pipe cracking is called hydrogen water chemistry (HWC). This paper presents a review of the results of EPRI's HWC development program from 1980 to the present. In addition, plans for additional work to investigate the feasibility of adapting HWC to protect the BWR vessel and major internal components from potential stress corrosion cracking problems are summarized.     

Electrochemical Corrosion Potential Modeling in the Primary Heat Transport Circuit of the Chinshan Boiling Water Reactor under the Condition of Hydrogen Water Chemistry with Noble Metal Coating

The technique of noble metal treatment, such as noble metal coating (NMC) or noble metal chemical addition, accompanied by a low level hydrogen water chemistry, is being employed by a number of nuclear power plants around the world for mitigating intergranular stress corrosion cracking in the vessel internals of their boiling water reactors (BWRs). A computer model DEM ACE was expanded and employed to assess the effectiveness of NMC throughout the primary heat transport circuit (PHTC) of a BWR. The effectiveness of NMC was justified by the electrochemical corrosion potential (ECP) and crack growth rate (CGR) predictions. In calculating the ECP, enhancing factors for the exchange current densities of redox reactions available from recently published data, were employed. The Chinshan BWR was selected as a model reactor. According to the modeling results, it was found that the effectiveness of NMC in the PHTC of a BWR could vary from region to region at different feedwater hydrogen concentrations. For the selected BWR, NMC was predicted to be of little benefit when the feedwater hydrogen concentration reached 0.9 ppm or over. In particular, the NMC technique proved to be beneficial in reducing ECP and CGR along the PHTC even if the BWR was operated under normal water chemistry.     

Water Chemistry in a Crack Tip under Irradiation, (II)

Under neutron and gamma-ray irradiations, radiolytic species are generated directly in the crack tip, which causes higher oxidant concentrations and subsequently influences crack propagation rate.

A crevice radiolysis model was proposed to estimate the oxidant concentrations in the crack tip water under gamma-ray irradiation. Direct generation of radiolytic species in the crevice water, and their secondary generation and disappearance caused by their interaction with the crevice surface as well as species in the crevice water were included in the model. The diffusion of the radiolytic species through the narrow gap from the bulk water to the crack tip and vice versa were also considered.

Calculation results confirmed that the concentrations of H₂O₂, one of the most important oxidants in BWR environments, in both bulk water and crack tip water under irradiation (energy deposition rate: 0.1 W/cm) were high enough to show high local ECP in both regions under NWC, but were high in the bulk water and low in the crack tip water under HWC. A high H₂ diffusion rate from the bulk to the crack tip enhanced the recombination reaction of H₂O₂ and H₂.     

Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors (IV)

A calculation model has been developed in order to evaluate effectiveness of hydrazine and hydrogen co-injection (HHC) into reactor water for mitigation of intergranular stress corrosion cracking of structural materials used in boiling water reactors (BWRs). The HHC uses the strong reducing power of hydrazine radical, which is produced in the downcomer region under irradiation by γ-rays and neutrons. Some reactions and their reaction rate constants were determined based on experiments which were carried out in aerated water, hydrogenated water, and deaerated water. The calculated results were in good agreement with experimental data by a factor of two. The model was applied to a BWR and it was found that the HHC cut oxygen and hydrogen peroxide amounts dissolved in reactor water more effectively than hydrogen water chemistry alone. Thus, the required amount of hydrogen for hydrazine injection was much lower than that for hydrogen water chemistry. Consequently, electrochemical corrosion potential of structural materials could be lowered below–0:1V vs. SHE without any increase of MS line dose rate, which has been a limitation of the conventional hydrogen water chemistry. The HHC was predicted to decrease crack growth rate of structural materials by a factor of 10.     

Effect of corrosion potential on the corrosion fatigue crack growth behaviour of low-alloy steels in high-temperature water

The low-frequency corrosion fatigue (CF) crack growth behaviour of different low-alloy reactor pressure vessel steels was characterized under simulated boiling water reactor conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in the temperature range of 240-288 °C with different loading parameters at different electrochemical corrosion potentials (ECPs). Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographical analysis by SEM were used to quantify the cracking response. In this paper the effect of ECP on the CF crack growth behaviour is discussed and compared with the crack growth model of General Electric (GE). The ECP mainly affected the transition from fast (‘high-sulphur’) to slow (‘low-sulphur’) CF crack growth, which appeared as critical frequencies ν_crit = f(ΔK, R, ECP) and ΔK-thresholds ΔK_EAC = f(ν, R, ECP) in the cycle-based form and as a critical air fatigue crack growth rate da/dt_Air,crit in the time-domain form. The critical crack growth rates, frequencies, and ΔK_EAC-thresholds were shifted to lower values with increasing ECP. The CF crack growth rates of all materials were conservatively covered by the ‘high-sulphur’ CF line of the GE-model for all investigated temperatures and frequencies. Under most system conditions, the model seems to reasonably well predict the experimentally observed parameter trends. Only under highly oxidizing conditions (ECP ? 0 mV_SHE) and slow strain rates/low loading frequencies the GE-model does not conservatively cover the experimentally gathered crack growth rate data. Based on the GE-model and the observed cracking behaviour a simple time-domain superposition-model could be used to develop improved reference CF crack growth curves for codes.     

Modeling the effects of nuclear environments on crevice corrosion of pure nickel by coupling of phase and polarization behavior

As from long-term operating experience the high purity primary water cycle of light water nuclear reactors may exhibit excursions from the recommended water chemistry leading to potentially favorite conditions for stress corrosion cracking (SCC) which may be initiated and its propagation controlled by local pitting and crevice corrosion. Deterministic modeling of local corrosion including incubation times for crevice corrosion should therefore provide a basis for lifetime predictions of components, which have been subjected to sporadic intermediate water chemistry fluctuations. Based on previous work for room temperature (RT), the chloride-induced crevice corrosion at 288 °C of pure nickel as an important base element in respective high alloyed nuclear materials is modeled by coupling anodic polarization with the precipitation of nickel oxide and nickel chloride calculated from the water–hydrogen–nickel chloride heterogeneous phase equilibrium diagram. The surface corrosion potentials are fixed by bulk levels of hydrogen and oxygen contents as well as pH simulating hydrogen treatment of irradiation subjected cooling water for the reduction of corrosion potentials and mitigation of SCC at operating temperature 288 °C in Boiling Water Reactors (BWRs). Assuming chemical equilibrium conditions during the selected time steps in a relevant component crevice the calculated change of the crevice solution composition is quantitatively shown to initiate crevice corrosion by the breakdown of the passive nickel oxide layer followed by the formation of non-passive nickel chloride and the subsequent acidification of the crevice solution. The effects of corrosion potentials, bulk levels of pH and chlorides, are investigated. As a result, the reduction of corrosion potentials and increase in bulk pH provide significant increases in the passive layer breakdown times and acidification times inside the crevice. Depending on bulk pH and corrosion potentials the reduction of bulk chlorides down to recommended levels in BWRs retards crevice corrosion significantly. For a standard 100,000 h time for crevice acidification to locally less than pH = 0 the respective chloride–pH domain is evaluated. Such diagrams may be related to respective effects on stress corrosion cracking and its mitigation by hydrogen water chemistry (HWC).     

Characterization of crud deposited on fuel rods under HWC environment in Kuosheng Nuclear Power Plant

Under normal water chemistry conditions, the oxygen and hydrogen peroxide produced by water radiolysis in the coolant of boiling water reactors(BWRs) can lead to intergranular stress corrosion cracking in the constituent materials of plant components. This fact has led to the wide-scale adoption of hydrogen water chemistry(HWC) in the nuclear industry to counteract these effects.This study seeks to characterize the metallic composition and the surface properties of the constituent materials of plant components in order to determine their effects on the accumulation of chalk river unidentified deposits(crud) on fuel rods in the BWR Unit-1 of the Kuosheng Nuclear Power Plant in Taiwan. Inductively coupled plasma-atomic emission spectroscopy was used to calculate the concentrations of surface crud and gamma spectrometry was used to determine the radioactivity of the corrosion products, as well as their axial distribution across the surface of the fuel rods. X-ray diffraction analysis and scanning electron microscopy/energy-dispersive X-ray spectroscopy were used to identify the crystalline phase and morphology of the crud as irregular shapes and flakes. The amount of crud deposited during the fourth fuel cycle exceeded that of the third fuel cycle due to extended burn-up time. Our analytical results indicate that the implementation of HWC had no significant effect on the characteristics of subsequent crud.     

The Efficiency of Noble Metals in Reducing the Corrosion Potential in the Primary Coolant Circuits of Boiling Water Reactors Operating under Hydrogen Water Chemistry Operation

In order to promote the effectiveness of hydrogen water chemistry (HWC) and to achieve a more effective reduction in electrochemical corrosion potential (ECP) in the primary coolant circuits of boiling water reactors (BWRs), the technology of noble metal chemical addition (NMCA) was brought into practice about 10 years ago. NMCA aims at enhancing the oxidation of hydrogen on metal surfaces and lowering the concentrations of the oxidants (oxygen and hydrogen peroxide) via recombination with hydrogen on the catalyzed surfaces, and therefore reducing the corrosion potentials of the structural alloys in a BWR primary heat transport circuit. Previous research indicates that the effectiveness of NMCA in combination with a low HWC might be evaluated via model predictions of the hydrogen-to-oxidant molar ratio (M_H/O) in the primary coolant circuit. If the M_H/O at a certain location is calculated to be greater than 2, it is justified that the NMCA would be effective in reducing the ECP to much below the critical potential for Intergranular Stress Corrosion Cracking (IGSCC), E_IGSCC, of --0.23 V_SHE. However, this statement is true only when the recombination efficiency of hydrogen with oxygen and/or hydrogen peroxide at the location of interest is 100%. Otherwise, significant amounts of oxidants may still be present, even with a stoichiometric M_H/O of greater than 2. With the aid of a computer model DEMACE, we explored the impact of incomplete recombination and found that the ECP might be reduced under given circumstances, but not to a great extent, and might remain well above E_IGSCC. Accordingly, considerable caution should be exercised upon using the M_H/O as a sole indicator for evaluating the effectiveness of NMCA with low HWC as a means of mitigating IGSCC in a BWR. An important finding of this study is that it is necessary to quantify the recombination efficiencies of hydrogen with oxygen and/or hydrogen peroxide on the noble metal treated stainless steel surfaces in order to qualify the use of M_H/O as an indicator for NMCA effectiveness in the primary coolant circuit of a BWR.     

Environmentally-assisted cracking behaviour in the transition region of an Alloy182/SA 508 Cl.2 dissimilar metal weld joint in simulated boiling water reactor normal water chemistry environment

The stress corrosion cracking (SCC) and corrosion fatigue behaviour perpendicular and parallel to the fusion line in the transition region between the Alloy 182 Nickel-base weld metal and the adjacent SA 508 Cl.2 low-alloy reactor pressure vessel (RPV) steel of a simulated dissimilar metal weld joint was investigated under boiling water reactor normal water chemistry conditions. A special emphasis was placed to the question whether a fast growing interdendritic SCC crack in the highly susceptible Alloy 182 weld metal can easily cross the fusion line and significantly propagate into the adjacent low-alloy RPV steel. Cessation of interdendritic SCC crack growth was observed in high-purity or sulphate-containing oxygenated water under constant or periodical partial unloading conditions for those parts of the crack front, which reached the fusion line. In chloride containing water, on the other hand, the interdendritic SCC crack in the Alloy 182 weld metal very easily crossed the fusion line and further propagated with a very high rate as a transgranular crack into the heat-affected zone and base metal of the adjacent low-alloy steel. The observed SCC cracking behaviour at the interface correlates excellently with the field experience of such dissimilar metal weld joints, where SCC cracking was usually confined to the Alloy 182 weld metal.     

Investigation of cobalt buildup behavior and suppression by zinc injection on stainless steel under HWC conditions using simultaneous continuous measurements of corrosion and cobalt buildup

In boiling water reactor (BWR) plants, cobalt-60 (⁶⁰Co) is the main source of radiation exposure, and it builds up on oxide films of structural materials. The ⁶⁰Co buildup is caused by its incorporation into the oxide films. In the BWR plants using hydrogen water chemistry (HWC) to mitigate the oxidative environment, Zn injection has been applied to reduce the ⁶⁰Co incorporation. In this work, we studied the incorporation mechanism of ⁶⁰Co into the oxide films on type 316 stainless steel and the suppression mechanism of ⁶⁰Co incorporation. In order to discriminate between coprecipitation and adsorption of ⁶⁰Co incorporation under HWC conditions, we measured the corrosion amount of the base metal and the ⁶⁰Co buildup amount, using simultaneous continuous measurements for 500 h. The ⁶⁰Co incorporation increased with time both with and without Zn injections. We found that the time dependencies of ⁶⁰Co incorporation with and without Zn have one and two regions, respectively. In the initial stage for both, ⁶⁰Co was incorporated mainly by coprecipitation. After 100 h without Zn, ⁶⁰Co was incorporated by both coprecipitation and adsorption. These results mean that Zn suppressed both coprecipitation and adsorption of ⁶⁰Co.