Hydrogen Water Chemistry Conditions Research Articles

Formulation of Irradiation Assisted Stress Corrosion Crack Growth Rates for Neutron-Irradiated Stainless Steels under Hydrogen Water Chemistry Conditions of Boiling Water Reactors

ABSTRACT This study developed a formula for the calculation of crack growth rates (CGRs) of irradiated austenitic stainless steels (SSs) under hydrogen water chemistry (HWC) conditions of boiling water reactors. First, the input parameters for the Hashimoto–Koshiishi model, which is a mechanistic crack growth model of stress corrosion cracking (SCC), were optimised for SSs irradiated up to 11 dpa. Electrochemical corrosion potential (ECP) of those SSs during crack growth tests nearly equalled −200 mVSHE. Second, a simplified formulation was developed to reproduce the model calculations of SSs. Calculated CGRs agreed well with experimental CGRs of various types of SSs, which were irradiated up to 13 dpa and tested at ECPs from −550 mVSHE to −200 mVSHE, when the experimental CGRs were less than the values for NWC conditions. On the other hand, there were some data in which experimental CGRs were much higher than calculated CGRs. By analysis of the experimental conditions of the data, the possible influencing factor for the discrepancy between the calculated and experimental CGRs was found to be the -validity. If the -validity criterion was not satisfied, the strain distribution used in the model calculation was not adequate for the actual strain distribution.

Impact of Chloride on the Environmentally-Assisted Crack Initiation Behaviour of Low-Alloy Steel under Boiling Water Reactor Conditions

Low-alloy reactor pressure vessel steels have a rather low susceptibility to stress corrosion cracking (SCC) in a boiling water reactor (BWR) environment if the high-temperature water contains no anionic impurities. Recent investigations revealed that under oxidizing BWR normal water chemistry (NWC) conditions extremely small amounts of chloride, can cause very high SCC growth rates in these materials. Therefore, the effect of continuous and temporary chloride additions on the crack initiation behaviour was explored by a series of constant extension rate tensile (CERT) and constant load tests in high-temperature water. In an NWC environment, containing ≥2 ppb of chloride, strain-induced corrosion cracking (SICC) initiation occurred briefly after the onset of plastic yielding and at much smaller strains than in high-purity water. On the other hand, under reducing hydrogen water chemistry conditions with up to 700 ppb chloride, no SICC was detected up to very high strains. CERT experiments, with moderate short-term chloride transients before and during the loading, showed that even serious mechanical loading transients, one day after returning to high-purity water, did not result in early SICC initiation.

Silicon dissolution and morphology modification of NITE SiC/SiC claddings in pressurized flowing water under neutron irradiation

Claddings of silicon carbide (SiC)-based composites for light water reactors are expected to be a key component of accident-tolerant fuel technology for enhancing nuclear safety. Muroran Institute of Technology has developed technologies for producing SiC/SiC claddings by a nano-infiltration and transient eutectic phase (NITE) method in the MEXT-funded project named “SCARLET.” As part of this program, neutron irradiation experiments were performed in a pressurized cooling water flow at the Halden Reactor in Norway. During the irradiation experiments, silicon dissolution from the SiC/SiC specimens was measured, and the relationship among the dissolution, properties of SiC/SiC claddings and reactor operating conditions was studied. The morphology of the SiC/SiC claddings was not significantly modified by the irradiation experiments; however, the joint parts were corroded and suffered damage. Observation of irradiated specimens and a rough estimate of the SiC dissolution rate under each irradiation condition indicated that the NITE SiC/SiC composite has potential as a structural material for pressurized water reactors (PWRs) that are operated under hydrogen water chemistry conditions. It is also shown that chemical-vapor-deposited (CVD)-SiC and other coating techniques are necessary for the realization of SiC reactor core. Joining and coating technologies were also tested, but appropriate methods could not be identified experimentally; thus, joining and coating techniques should be further developed.

Stress corrosion cracking initiation and short crack growth behaviour in Alloy 182 weld metal under simulated boiling water reactor hydrogen water chemistry conditions

The effect of dissolved hydrogen on the stress corrosion cracking initiation and short crack growth behaviour of Alloy 182 weld metal was evaluated in 274°C hydrogenated high-purity water using accelerated crack initiation and growth tests with sharply notched fracture mechanics specimens and in-situ crack initiation and growth monitoring. A maximum in initiation susceptibility and crack growth rates was observed at the Ni/NiO phase transition line. Grain boundary misorientations and mismatch in Schmid factor were measured by electron backscattered diffraction along intergranular stress corrosion cracks and at crack initiation sites. Low-angle boundaries seem to be particularly resistant to cracking.

Development of a method to lower recontamination after chemical decontamination by depositing Pt nano particles: (II) consideration of the Pt effect on oxide composition

ABSTRACTThe Pt coating (Pt-C) process has been developed to lower the recontamination by 60Co which was incorporated in oxides on piping surface after chemical decontamination. In order to determine the suppression mechanism of 60Co deposition by Pt-C, it is important to investigate the formation of oxide film 60Co deposition behavior on oxide with Pt-C specimens. In this paper, we observed the composition change of oxide after a 60Co deposition test under the hydrogen water chemistry condition, and considered the 60Co deposition behavior on oxide for Pt-C specimens. The Ni and Co metal concentrations in oxide were dramatically changed by Pt-C process. The Ni metal concentrations in oxide for specimens with and without the Pt-C process were 11.2% and 18.0%, respectively. On the other hand, the Co metal concentrations in oxide for specimens with and without the Pt-C process were 1.2% and 0.2%, respectively. This composition change of the oxides indicated that 60Co incorporation for the Pt-C specimens was suppressed by replacing 60Co with Ni. We concluded that the Ni2+ ions were incorporated into the 8a site of the oxide spinel structure instead of Co2+ ions due to the effect of the conversion deposition energy.

Uniform corrosion of FeCrAl alloys in LWR coolant environments

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300–500 μm thick cladding.

Oxide Film Characterization after the crack propagation in CT specimens of AISI 304L under Hydrogen Water Chemistry Condition.

ABSTRACTStress Corrosion Cracking (SCC) in a general term describing stressed alloy fracture that occurs by crack propagation in specifically environments, and has the appearance of brittle fracture, yet it can occur in ductile materials like AISI 304L used in internal components of Boiling Water Reactors (BWR). The high levels of oxygen and hydrogen peroxide generated during an operational Normal Water Condition (NWC) promotes an Electrochemical Corrosion Potential (ECP), enough to generate SCC in susceptible materials. Changes in water chemistry have been some of the main solutions for mitigate this degradation mechanism, and one of these changes is reducing the ECP by the injection of Hydrogen in the feed water of the reactor; this addition moves the ECP below a threshold value, under which the SCC is mitigated (-230mV vs SHE). This paper shows the characterization by Scanning Electron Microscopy (SEM), X-Ray Diffraction (XRD) and Raman Spectroscopy of the oxide film formed in to a crack propagated during a Rising Displacement Test method (RDT), on Hydrogen Water Chemistry (HWC) conditions: 20 ppb O2, 125 ppb H2, P=8MPa, T=288°C, using a CT specimen of austenitic stainless steel AISI 304L sensitized. The characterization allowed identifying the magnetite formation since an incipient way, until very good formed magnetite crystals.

Cobalt Radioactivity Behaviors in a BWR Environment and Countermeasures for Dose Rate Reduction

While under normal water chemistry without any specific metal ions in reactor coolant a high electrochemical corrosion potential caused by highly oxidizing species such as hydrogen peroxide promotes the formation of hematite film on piping surfaces with a densely packed film structure, the presence of a certain amount of nickel ions prevents the magnetite film from changing to hematite by forming a nickel ferrite. This formation of nickel ferrite instead of hematite accelerates cobalt buildup, and this is especially notable for carbon steel. The observed reduction of radioactivity concentration in reactor water by zinc injection or by nickel/iron ratio control can be explained by the role of zinc or nickel in preventing the film on the fuel rod surfaces from changing to hematite, thereby stabilizing the cobalt activity on this surface. A thermodynamic evaluation suggests that zinc ferrite is more stable than cobalt ferrite only when the ratio of cobalt to zinc divalent ions, [Co2+]/[Zn2+], is <0.011 in molar units. This ratio is consistent with the ratio of 60Co activity to zinc concentration commonly used in industry to control reactor water zinc levels for a dose rate reduction under the hydrogen water chemistry condition. Based on the present understanding of radioactivity behaviors, the actual radiation dose reduction methods are classified into the several groups and summarized from the viewpoint of the interaction between the oxide and various metal ions.

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.

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.

Effect of Pt injection rate on corrosion potential and Pt distribution on stainless steel under simulated boiling water reactor conditions

Online NobleChem is a technology to mitigate stress corrosion cracking in reactor internals and recirculation pipes of boiling water reactors (BWRs). For a more efficient reduction of the electrochemical corrosion potential (ECP) under hydrogen water chemistry conditions, noble metals (e.g. Pt) are injected into the feed water during power operation. They are claimed to deposit as very fine metallic particles on all water wetted surfaces and to stay electrocatalytic over long periods. To study the deposition and (re)distribution behaviour of Pt, a sophisticated high temperature water loop has been constructed, in which specimens can be exposed to simulated BWR water and Pt can be deposited in situ. The specimens are analysed by laser ablation–inductively coupled plasma–mass spectrometry and electron microscopy after the experiments. Results from two experiments with different Pt injection rates but similar total Pt amounts revealed a faster and more pronounced reduction of the ECP and slightly larger average Pt particle size using the higher Pt injection rate. A longer preoxidation of the specimens seems to increase the average Pt concentration on the specimen surface.

Corrosion fatigue crack growth behaviour of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The corrosion fatigue crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was systematically characterized under simulated boiling water reactor normal water and hydrogen water chemistry conditions by low-frequency fatigue 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 this paper, the observed synergistic effects of environmental, material and loading parameters on the environmental acceleration of fatigue crack growth in low-alloy RPV steels are discussed in the context of the Ford–Andresen model. Additionally, the adequacy and conservatism of the current “ASME XI reference fatigue crack growth curves” of the ASME Boiler & Pressure Vessel Code are critically reviewed and assessed on the basis of the gathered experimental data base and this model. Based on the observed cracking behaviour and the Ford–Andresen model, a simple time-domain superposition model is suggested, which could reduce most of the undue conservatism and eliminate uncertainties of the existing codes and therefore serve as a basis for the development of improved reference fatigue crack growth curves.

Effects of Hydrogen Peroxide on Cobalt Deposition Rate on Primary Cooling System ofBoiling Water Reactors

Experimental data of radioactivity deposition rates on stainless steel specimensexposed to high-temperature water including hydrogen peroxide (H2O2) and oxygen (O2) were applied to the evaluation of Co-60 accumulation rates in oxide films on the stainless steel piping of a boiling water reactor (BWR) primary cooling system and then an empirical formula of Co-60 deposition rate coefficient was proposed as a function of exposure time. The empirical formula was applied to the evaluationof the amount of Co-60 deposition on the stainless steel piping of BWR primary cooling systems under normal water chemistry (NWC) and hydrogen water chemistry (HWC) conditions. The results are summarized as follows. 1) The Co-60 deposition rate coefficient was almost independent of O2 concentration ([O2]), while it decreased as H2O2 concentration ([H2O2]) decreased. 2) The proposed empirical formula determined on the basis of experiments under H2O2 conditions could give a much better estimation of the Co-60 deposition rate coefficient on stainless steel under NWC conditions of BWR than the previous formula based on the data under simple O2 conditions. 3) The deposition amounts of Co-60 on stainless steel piping under HWC conditions were evaluated using the formula, which suggested that those without any prefilming treatment were several times as high as those under NWC conditions. 4) The increase in Co-60 deposition under HWC conditions was caused by the less protective film in mitigating Co-60 deposition and it could be avoided by several hundred hours of operation under NWC conditions prior to HWC operation. The effect on material integrity is evaluated to be negligible.

Fatigue Crack Growth Behavior of Sensitized Type 304 Stainless Steel under Boiling Water Reactor Conditions

Abstract The fatigue crack growth behavior of sensitized Type 304 (UNS S30400) stainless steel (SS) with an electrochemical potentiokinetic reactivation (EPR) value of 30 C/cm2 was investigated under boiling water reactor (BWR) conditions at 288°C. Tests were performed under either cyclic or trapezoidal loading conditions. Test environments were high-purity water either with 200 ppb dissolved oxygen (DO) under normal water chemistry (NWC) conditions or with 10 ppb DO plus 160 ppb dissolved hydrogen (DH) under hydrogen water chemistry (HWC) conditions. It was observed that whether the loading waveform was cyclic or trapezoidal, the fatigue crack growth rates (FCGR) could not be reduced by HWC conditions. None of the current FCGR data was encompassed by the reference curves of the ASME Code Section XI, Appendix C (ASME-XI, Appendix C), whereas all of the current data could be bounded by the JSME flaw evaluation code (JSME S NA1-2000). Finally, the examination of the accuracy of the superposition principles ...

Analysis of water radiolysis in relation to stress corrosion cracking of stainless steel at high temperatures – Effect of water radiolysis on limiting current densities of anodic and cathodic reactions under irradiation

Electrochemical corrosion potential (ECP) is an important measure for environmental factor in relation to stress corrosion cracking (SCC) of metal materials. In the case of SCC for in-core materials in nuclear reactors, radiolysis of coolant water decisively controls ECP of metal materials under irradiation. In the previous models for ECP evaluation of stainless steel, radiolysis of reactor water in bulk was considered to calculate the bulk concentrations of the radiolysis products. In this work, the radiolysis not only in bulk but also in the diffusion layer at the interface between stainless steel and bulk water was taken into account in the evaluation of ECP. The calculation results shows that the radiolysis in the diffusion layer give significant effects on the limiting current densities of the redox reactions of the radiolysis products, H 2O 2 and H 2, depending on dose rate, flow rate and water chemistry, and leads to the significant increase in the ECP values in some cases, especially in hydrogen water chemistry conditions.

Effects of Noble Metal Deposition upon Corrosion Behavior of Structural Materials in Nuclear Power Plants, (I)

The effects of noble metal deposition under hydrogen water chemistry (HWC) condition on the features of oxide film formed on structural components in a reactor were studied. Noble metal-deposited type 304 stainless steel specimens with an oxide film were exposed to the simulated HWC condition, including co-existing Co radioactivity. Relationships between features of the oxide film which had two layers and the accumulation and distribution of Co radioactivity in the oxide film were established. The outer layer of the oxide film which consisted of α-Fe2O3, Fe3O4 and NiFe2O4 was dissolved by noble metal deposition and exposure to the HWC condition. The reasons for this were as follows. Solubility of α-Fe2O3, Fe3O4 and NiFe2O4 increased with the decrease of electrochemical corrosion potential. Dissolution of these compounds was accelerated by the anodic reaction of hydrogen which is catalyzed by noble metal. Co radioactivity was mainly incorporated into the inner layer. This was caused by the substitution of radioactive Co ions for ferrous ions in the oxide film, based on the observation that growth of the oxide film and oxidation of base metal stopped in the HWC condition. The inner layer consisted of FeCr2O4 which is stable at low ECP and it did not dissolve. Co radioactivity was not incorporated into the outer layer because it dissolved.

Effects of Noble Metal Deposition upon Corrosion Behavior of Structural Materials in Nuclear Power Plants, (I): Effect of Noble Metal Deposition with an Oxide Film on Type 304 Stainless Steel under Simulated Hydrogen Water Chemistry Condition

The effects of noble metal deposition under hydrogen water chemistry (HWC) condition on the features of oxide film formed on structural components in a reactor were studied. Noble metal-deposited type 304 stainless steel specimens with an oxide film were exposed to the simulated HWC condition, including co-existing Co radioactivity. Relationships between features of the oxide film which had two layers and the accumulation and distribution of Co radioactivity in the oxide film were established. The outer layer of the oxide film which consisted of α-Fe2O3, Fe3O4 and NiFe2O4 was dissolved by noble metal deposition and exposure to the HWC condition. The reasons for this were as follows. Solubility of α-Fe2O3, Fe3O4 and NiFe2O4 increased with the decrease of electrochemical corrosion potential. Dissolution of these compounds was accelerated by the anodic reaction of hydrogen which is catalyzed by noble metal. Co radioactivity was mainly incorporated into the inner layer. This was caused by the substitution of radioactive Co ions for ferrous ions in the oxide film, based on the observation that growth of the oxide film and oxidation of base metal stopped in the HWC condition. The inner layer consisted of FeCr2O4 which is stable at low ECP and it did not dissolve. Co radioactivity was not incorporated into the outer layer because it dissolved.

Investigation of Cobalt Deposition Behavior with Zinc Injection on Stainless Steel under BWR Conditions

Radioactive58Co and 65Zn were used together to investigate the deposition behavior of cobalt and zinc in the oxide film formed on the stainless steel piping of an apparatus simulating the high temperature and pressure conditions of a boiling water reactor. Using these radioactive tracers allowed the experimental setup to closely approximate the conditions found in actual plants. The accumulation of 58Co and 65Zn on the stainless steel piping was monitored using an online gamma-ray detector. The results were as follows. At higher zinc concentrations, the early period of fast deposition was shortened and the later deposition phase was slowed. Two mechanisms appeared to be responsible for the suppression of cobalt deposition by zinc injection. The first worked by decreasing the growth rate of the oxide film. This effect appeared most conspicuously in the outer oxide layer under normal water chemistry conditions. The second mechanism worked by reducing the cobalt concentration in the oxide film. This second effect was prominent in the inner oxide layer under hydrogen water chemistry conditions.

Effect of Corrosion Potential and Microstructure on the Stress Corrosion Cracking Susceptibility of Nickel-Base Alloys in High-Temperature Water

Abstract To evaluate the effect of corrosion potential on the stress corrosion cracking (SCC) initiation lifetime of nickel-base alloys, periodically unloaded uniaxial constant load (PU-UCL) tests were conducted, simulating boiling water reactor (BWR) water chemistry conditions. Utilizing the exponential distribution model for the determination of SCC lifetime, the SCC improvement factor was assessed by comparing the statistical location parameter of SCC failure time in PU-UCL tests between the hydrogen water chemistry (HWC) condition and the normal water chemistry (NWC) condition. The improvement factor larger than 10 was obtained for the moderate HWC condition, whereas it became more than 30 under the same condition when noble metals were chemically doped on the specimen surface preliminarily. The effect of microstructure on the SCC susceptibility was discussed after creviced bent beam (CBB) tests and microstructural analysis on grain boundaries. Alloy 182 (UNS W86182) with dendrite structure exhibits n...

Precise Evaluation of Corrosion Environments of Structural Materials under Complex Water Flow Condition, (II)

In order to clarify the corrosion environment of nuclear reactor structure material precisely, the water radiolysis evaluation considering a flow of reactor water was performed. This study focused on the downcomer region of a Boiling Water Reactor (BWR). Radiation intensity of the region is high and water radiolysis reaction occurs well. Therefore, the region is recognized as important for recombination of oxygen, hydrogen peroxide and hydrogen back to water under hydrogen water chemistry (HWC) condition. Precise evaluation of water chemistry in the region with consideration of water flow, however, has not been performed. In the downcomer region water flow is very complicated because the jet pumps are close together and water flows out to the recirculation system through only two nozzles. A new water radiolysis model was developed which can simulate water radiolysis under the three-dimensional water flow condition. The calculation was performed using the result of water flow analysis obtained by the computer code “a-Flow.” The water chemistry near a shroud wall and the reactor pressure vessel (RPV) wall differed greatly, and especially so under HWC condition. Also, the detailed distribution of the electrochemical corrosion potential (ECP) of structural materials was clarified by using the water chemistry analysis result.