Primary Coolant Of Pressurized Water Reactors Research Articles

Corrosion of 316L stainless steel produced by laser powder bed fusion and powder metallurgy in pressurized water reactor primary coolant

Corrosion mechanism of 316 L stainless steel produced by laser powder bed fusion-hot isostatic pressing (LPBF-HIP) and powder metallurgy-hot isostatic pressing (PM-HIP) is studied with in-situ electrochemical impedance measurements coupled to detailed oxide film characterization. Quantitative analysis of impedance spectra using the Mixed-Conduction Model and estimation of local kinetic and transport parameters by interpretation of in-depth elemental composition profiles indicated lower corrosion and oxidation rates of LPBF-HIP and PM-HIP materials in comparison to conventional wrought 316 L. This owes to a higher fraction of low-angle grain boundaries, smaller grain size, the presence of nano-sized oxide particles and elevated Cr and Ni contents.

Oxidation of Alloy 690 in Simulated Pressurized Water Reactor Primary Coolant – Experiments and Modeling

During normal operation of the primary loop of a water cooled-water moderated energy reactor (WWER), the associated water chemistry evolves from an initially high-boron high-potassium (beginning-of-cycle, 1200 ppm B and 8 ppm K) to significantly lower concentrations (ca. 400 ppm B and 3.5 ppm K) at the end of the respective campaign. In addition to that, the concentration of Li produced by transmutation of B reaches ca. 0.7 – 0.9 ppm. To the best of our knowledge, the effect of such water chemistry changes on pressurized water reactor (PWR) internals such as Alloy 690 have not been studied yet. Such a study appears relevant also within the frames of a discussion for a possible transition from B-Li to B-K water chemistry in PWRs.In the present paper, the effect of evolution of primary water chemistry during operation on the corrosion rate and conduction mechanism of oxide films on Alloy 690 is studied by in-situ impedance spectroscopy at 280 °C / 8 MPa during 1-week exposures in an autoclave connected to a re-circulation loop. Impedance measurements were performed in three-electrode mode featuring a Pt sheet counter electrode and a Pd electrode negatively polarized with a current of -10 µA vs. another Pt to approximate the reversible hydrogen electrode (RHE). Spectra were recorded at the corrosion potential in a frequency range 10 kHz – 0.1 mHz using an ac amplitude of 50 mV (rms). At the end of exposure, the samples were polarized in the range of potentials -0.1 to 0.5 V vs. RHE to evaluate the stability of the passive oxide. Separate samples of Alloy 690 were simultaneously exposed in the autoclave and subsequently analyzed by Glow-Discharge Optical Emission Spectroscopy (GDOES) in order to estimate the thickness and the in-depth composition of the formed oxides.Impedance data were quantitatively interpreted using the Mixed-Conduction Model for Oxide Films (MCM) to estimate the rates of metal oxidation at the alloy/oxide interface, oxide dissolution and restructuring at the film/coolant interface and ion transport in the protective layer. The effect of water chemistry evolution on corrosion rate and conduction mechanism in the oxide on Alloy 690 in primary coolant is discussed on the basis of the obtained parameters.Acknowledgments: This study is funded by the European Union-NextGenerationEU, through the National Recovery and Resilience Plan of the Republic of Bulgaria, project № BG-RRP-2.004-0002 (BiOrgaMCT).

Optimization of dissolved hydrogen concentration for mitigating corrosive conditions of pressurized water reactor primary coolant under irradiation (1) evaluation of water radiolysis

ABSTRACT A major subject for evaluating the corrosive conditions in the pressurized water reactor (PWR) primary coolant is to determine the optimal hydrogen concentration for mitigating primary water stress corrosion cracking (PWSCC) without adverse effects on major structural materials. An analytical method combining water radiolysis and electrochemical corrosion potential (ECP) analyses was proposed for evaluating the corrosive conditions in PWR primary coolant. These procedures originally developed for boiling water reactors (BWRs) were successfully extended to PWRs with different water chemistry parameters, including pH, temperature, and radiation qualities, after minor changes in the original procedures and major input parameters, such as the inclusion of the effects of alpha radiolysis and Li+ (Na+) and H+ effects for the anodic polarization curve. This study discusses the results of water radiolysis analysis for PWR primary coolant, and the characteristic behavior of hydrogen peroxide (H2O2) as a function of hydrogen inlet concentrations in PWR primary coolant conditions (higher pH and α-ray irradiation). A possible reaction scheme involving eaq – was proposed for H2O2 suppression under alkaline conditions. The corrosive conditions were discussed in the following publication, using ECP as the major index for PWR corrosive conditions.

Optimization of dissolved hydrogen concentration for mitigating corrosive conditions of pressurised water reactor primary coolant under irradiation (2) evaluation of electrochemical corrosion potential

ABSTRACT One of the major subjects for evaluating the corrosive conditions in the PWR primary coolant was to determine the optimal hydrogen concentration for mitigating PWSCC without any adverse effects on major structural materials. As suitable procedures for evaluating the corrosive conditions in PWR primary coolant, a couple of procedures, i.e. water radiolysis and ECP analyses, were proposed. The previous article showed the radiolysis calculation in the PWR primary coolant, which was followed by an ECP study here. The ECP analysis based on a couple of a mixed potential model and an oxide layer growth model was developed originally for BWR conditions, which was extended to PWR conditions with adding Li+ (Na+) and H+ effects on the anodic polarization curves. As a result of comparison of the calculated results with INCA in-pile-loop experiment data as well as other experimental data, it was confirmed that the ECPs calculated with the coupled analyses agreed with the measured within ±100 mV discrepancies.

Some fundamental understandings of Zn-injection water chemistry on material corrosion in pressurized water reactor primary circuit

The optimization of water chemistry in pressurized water reactor (PWR) primary circuit is one of the most effective ways to achieve both safety and economy for operating PWR nuclear power plants (NPPs). Special attention has been paid to fundamental research and engineering application of Zn-injection water chemistry (ZWC) in PWR primary circuit in recent years in China due to the rapid development of nuclear power. The present paper mainly reports the status of PWR NPPs in China and some fundamental understandings of ZWC on material corrosion. Effects of temperature (T), pH at T value, Zn concentration, Zn-injection sequence and lasting time on electrochemical corrosion behavior, oxide film characteristic and stress corrosion cracking (SCC) susceptibility of nuclear-grade austenitic alloys were investigated. A modified point defect model was used to discuss the effects of ZWC on the oxide films in high-temperature pressurized water. Some water chemistry parameters in PWR primary coolant system are proposed for mitigating corrosion and SCC of nuclear-grade austenitic alloys.

Three Loading Patterns for 99Tc Transmutation in PWR Burnable Poison Rods

Using the MCNP code, we explore three different 99Tc loading patterns in pressurized water reactor (PWR) burnable poison rods (BPRs). We also calculate the effects on the PWR keff and boric acid concentration readjusted amount after loading 99Tc transmutation material into PWR BPRs. Finally, we carry out the transmutation rate and depletion calculation.After the 99Tc transmutation material is mixed homogeneously with burnable poison (BP) in the BPRs, keff slightly increases. As the amount of 99Tc coating on the BPRs increases, keff decreases gradually and slightly; this result is similar to the tendency of keff to decrease after applying a thin-layer coating of minor actinide in the BPRs. Our calculation results show that as the coating thickness of 99Tc in the water gap of BPRs increases, keff decreases correspondingly. The more BPR water gaps are filled in with 99Tc transmutation material, the sharper is the decrease of keff. If 99Tc fills in the water gaps of 12 BPRs of each fuel assembly, the coating thickness is 0.02 cm, and the corresponding total 99Tc coating amount is 206.76 kg. This is the annual 99Tc yield of more than three PWRs.Our calculations also indicate that the more 99Tc is loaded into the PWR, the more boric acid concentration needs to be reduced in the coolant. For instance, if the 99Tc coating thickness in the water gaps is 0.02 cm, when 99Tc fills in 12 BPR water gaps of each fuel assembly, a boric acid concentration of 75 parts per million must be reduced from the PWR primary coolant to allow the PWR return to criticality. The transmutation rate and burnup calculation results indicate that 99Tc mixed homogeneously with BP may be a satisfactory 99Tc loading pattern in 99Tc transmutation in PWRs.

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.

Corrosion of nickel-base alloys in primary medium of pressurized water reactors: New insights on the oxide growth mechanisms and kinetic modelling

The generalised corrosion of Ni-base alloys exposed to pressurized water reactors primary coolant leads both to the growth of a protective oxide scale and to the release of cations. Based on experimental results previously published, a mechanistic description of this phenomenon is proposed. Then, from a simplified view of theses mechanisms, a kinetic model is proposed, focusing on the link between the alloy subsurface defects density and the shape of oxidation kinetic curves. Comparing calculated curves and experimental data shows this first approach is promising. Nevertheless a better knowledge of some kinetic parameters is necessary to improve the model.

환원제염조건에서 가압경수로 구조재료의 틈부식 특성

Crevice corrosion tests were conducted to examine the corrosion properties of HYBRID (HYdrazine Base Reductive metal Ion Decontamination) which was developed to decontaminate the PWR primary coolant system. To compare the corrosion properties of HYBRID with commonly existing decontamination agents, oxalic acid (OA) and citric oxalic acid (CITROX) were also examined. Type 304 Stainless Steel (304 SS) and Alloy 600 which are major components of the primary coolant system in Pressurized Water Reactor (PWR) were evaluated. Crevice corrosion tests were conducted under very aggressive conditions to confirm quickly the corrosion properties of primary coolant system structure components which have high corrosion resistance. Pitting and IGA were occurred in crevice surface under OA and CITROX conditions. But localized corrosion was not observed under HYBRID condition. Very low corrosion rate of less than 1.3×10-3 μm/h was observed under HYBRID condition for both materials. On the other hand, under OA condition, Alloy 600 indicated comparatively uniform corrosion rate of 4.0×10-2 μm/h but 304 SS indicated rapid accelerated corrosion in lower case than pH 2.0. In case of HYBRID condition, general corrosion and crevice corrosion were scarcely occurred. Therefore, material integrity of HYBRID in decontamination of primary coolant system in pressurized water reactor (PWR) reactor was conformed.

DETERMINATION OF THE 129I IN PRIMARY COOLANT OF PWR

Among the radioactive wastes generated from the nuclear power plant, a radioactive nuclide such as 129 I is classified as a difficult-to-measure (DTM) nuclide, owing to its low specific activity. Therefore, the establishment of an analytical procedure, including a chemical separation for 129 I as a representative DTM, becomes essential. In this report, the adsorption and recovery rate were measured by adding 125 I as a radio-isotopic tracer (t 1/2 = 60.14 d) to the simulation sample, in order to measure the activity concentration of 129 I in a pressurized-water reactor primary coolant. The optimum condition for the maximum recovery yield of iodine on the anion exchange resins (AG1 x2, 50-100 mesh, Cl − form) was found to be at pH 7. In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of 129 I was examined, as was the effect of 3 H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous 3 H presence was found with activity concentrations of 3 H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 μg/mL.

The Point Defect Model for Bi-Layer Passive Films

The Point Defect Model (PDM) has been shown to accurately describe the properties of passive films that form on metal surfaces in contact with aggressive environments under both open circuit and anodic polarization conditions. However, the commonly-employed PDM, known henceforth as PDM-II assumes that passivity arises from the properties of the barrier layer and that the outer layer, if present, contributes negligibly to the interfacial impedance. In this paper, we describe PDM-III, in which a resistive outer layer exists on the surface and contributes substantially to the impedance of the interface. The outer layer is shown to have a profound impact on the properties of the barrier layer and under certain circumstances the barrier layer is predicted to disappear. This new form of depassivation has been observed experimentally. The use of electrochemical impedance spectroscopy to characterize passive films having resistive outer layers is describe and illustrated with reference to the passive state on zirconium in simulated PWR (Pressurized Water Reactor) primary coolant.

Photoelectrochemical study of nickel base alloys oxide films formed at high temperature and high pressure water

The oxide film formed on nickel base alloys at high temperature and high pressure water exhibits semi-conducting properties evidenced by photocurrent generation when exposed to monochromatic light. The use of macro- and micro-photoelectrochemical techniques (PEC and MPEC) aims to identify the different semiconductor phases and their distribution in the oxide film. Three different nickel base alloys were corroded in recirculation loop at 325 °C in pressurised water reactor primary coolant conditions for different exposition durations. PEC experiments on these materials enable to obtain macroscopic energy spectra showing three contributions. The first one, with a band gap around 2.2 eV, was attributed to the presence of nickel hydroxide and/or nickel ferrite. The second one, with a band gap around 3.5 eV, was attributed to Cr2O3. The last contribution, with a band gap in the range of 4.1–4.5 eV, was attributed to the spinel phase Ni1−xFexCr2O4. In addition, macroscopic potential spectra recorded at different energies highlight n-type semi-conduction behaviours for both oxides, Cr2O3 and Ni1−xFexCr2O4. Moreover, MPEC images recorded at different energies exhibit contrasted regions in photocurrent, describing the distribution of nickel hydroxide and/or nickel ferrite and Cr2O3 in the oxide film at a micron scale. It is concluded that PEC techniques represent a sensitive and powerful way to locally analyse the various semiconductor phases in the oxide scale.

Measurement of 90Sr in primary coolant of pressurized water reactor

The present procedure for analysing 90Sr combines the use of a non radioactive carrier with high pressure ion chromatography separation, conductivity detection of the carrier and optimized external counting by liquid scintillation. This improvement with respect to traditional methods led to a more rapid and efficient purification stage. The present work proves that activities of 90Sr as low as 3 Bq/L can be measured in highly contaminated pressurized water reactor primary coolant matrix without any observed radiochemical interference. The approach shows promise for the analysis of other emitters of low energy radiation, or isotopes subject to high background or matrix effects in a PWR primary coolant.

Characterization of the Oxide Films Formed at the Surface of Ni-Base Alloys in Pressurized Water Reactors Primary Coolant by Transmission Electron Microscopy

The oxide film formed on nickel-based alloys in Pressurized Water Reactors (PWR) primary coolant conditions (325°C, aqueous media) has been investigated by Transmission Electron Microscopy (TEM). TEM observations revealed an oxide layer divided in two parts. The internal layer was mainly composed of a continuous spinel layer, identified as a mixed iron and nickel chromite (Ni(1-x)FexCr2O4). Moreover, nodules of Cr2O3 were present at the interface between this spinel and the alloy. The external layer is composed of large crystallites corresponding to a spinel structure rich in iron (Ni(1-z)Fe(2+z)O4) resulting from precipitation phenomena. The influence of alloy surface defects was also studied underlining two main consequences on the formation of the passive film e.g. the internal layer. On one hand, the growth kinetics of the internal spinel rich in chromium increased with the surface defect density. Besides that, when the defect density increased, the oxide scale became more finely crystallized. This result agrees with a growth mechanism due to a rate limiting process of diffusion through the grain boundaries of the oxide. On the other hand, the quantity of Cr2O3 nodules increased with the number of surface defects, revealing that the nodules nucleated preferentially at defect location.

The impact of radiolytic yield on the calculated ECP in PWR primary coolant circuits

A code, PWR–ECP, comprising chemistry, radiolysis, and mixed potential models has been developed to calculate radiolytic species concentrations and the corrosion potential of structural components at closely spaced points around the primary coolant circuits of pressurized water reactors (PWRs). The pH( T) of the coolant is calculated at each point of the primary-loop using a chemistry model for the B(OH) 3 + LiOH system. Although the chemistry/radiolysis/mixed potential code has the ability to calculate the transient reactor response, only the reactor steady state condition (normal operation) is discussed in this paper. The radiolysis model is a modified version of the code previously developed by Macdonald and coworkers to model the radiochemistry and corrosion properties of boiling water reactor primary coolant circuits. In the present work, the PWR–ECP code is used to explore the sensitivity of the calculated electrochemical corrosion potential (ECP) to the set of radiolytic yield data adopted; in this case, one set had been developed from ambient temperature experiments and another set reported elevated temperatures data. The calculations show that the calculated ECP is sensitive to the adopted values for the radiolytic yields.

Effect of post weld heat treatment on stress corrosion cracking of a low alloy steel to stainless steel transition weld

The susceptibility to stress corrosion cracking (SCC) of a low alloy to stainless steel transition weld, A508-309L/308L, in simulated pressurised water reactor (PWR) primary water, contaminated with low levels of sulphate, at 292°C was assessed via slow strain rate tests under controlled potentials and static load tests at free corrosion potentials. The minimum potential for SCC, E c, and crack growth rate were used to evaluate the susceptibility of as-welded and post-weld heat treated material. U-bend specimens tested at open circuit potential showed no SCC. The results indicate that if exposed, the transition weld should be immune to SCC in PWR primary side coolant water.

Some aspects of Ni behaviour in PWR primary coolant

Nickel chemistry in a pressurised water reactor (PWR) primary circuit is important both because it can be activated to 58 Co and because it is a major constituent of fuel clad deposits in plants affected by axial offset anomalies (AOAs). It has been suggested that a small decrease in the coolant dissolved H 2 concentration would switch the stable species from metallic nickel to NiO, which would deposit less as it is more soluble and thus reduce the extent of activation. Furthermore, it has been argued that the presence of NiO rather than metallic nickel in AOA deposits is an indication of oxidising conditions in the core. To understand nickel chemistry in a PWR core, it is necessary to consider not only thermodynamics but also radiation chemistry effects. In addition, some thermal processes may be kinetically constrained. Consequently, a thermodynamic model has been developed to simulate nickel chemistry. This paper discusses the results of thermodynamic and kinetic modelling and suggests that further experimental data are needed.

Stress corrosion cracking of a low alloy steel to stainless steel transition weld in PWR primary waters at 292°C

The effects of electrode potential and sulphate ( SO 2− 4 ) content in the water on the stress corrosion cracking (SCC) of a low alloy steel — austenitic stainless steel transition weld, A508-309L/308L, in pressurised water reactor (PWR) primary side waters at 292°C have been studied using slow strain rate testing (SSRT). The weld was post-weld-heat-treated at 620°C for 20 h before testing. Results showed that the transition zone in the weld had a higher susceptibility to SCC than either the bulk stainless steel or the bulk low alloy steel. The SCC in the transition zone was mainly intergranular in the austenitic layer, but transgranular cracking occurred at the interface and in the low alloy steel. The minimum potential for SCC, E c , in each water used was higher than the free corrosion potential range of −880 to −660 mV (SHE) and the susceptibility to SCC increased with increasing electrode potential. In sulphate doped waters, crack growth rates >2 × 10 −6 mm/s occurred at high applied potentials in the low alloy steel and/or in the austenitic layer but some less severe cracking occurred at the interface. Contamination of the water with SO 2− 4 increased the SCC susceptibility by both decreasing the minimum potentials for cracking and increasing the crack growth rate. However, the data suggest that transition welds should be immune from SCC in typical PWR primary side coolant water at 292°C even in the unlikely event that a break in the stainless steel cladding allowed access of the cooling water to the transition joint area.

A limit load criterion to predict crack growth in stainless steel pipes

In a recent test program, specimens of circumferentially cracked type 304 stainless steel pipes were subjected to dynamic cyclic loading [E. Murakami et al., in Advances in Fracture and Fatigue for the 1990 s (Edited by G. Yagawa and G. M. Wilkowski), Vol. 4, pp. 115–121 (1989)]. The experimental data indicated a linear correlation between the limit load of the pipe's cross-section, assuming elastic-plastic material behavior, and the logarithm of the number of loading cycles which are required to drive the crack through the pipe's thickness. A similar criterion is postulated to investigate the crack growth behavior observed in a High Level Vibration Test (HLVT) Program performed on a large scale modified model of a pressurized water reactor primary coolant system made of an equivalent stainless material [Y. J. Park et al., Final Report, NUREG/CR-5585 (May 1991)]. The input motion in the HLVT Program induced inelastic stresses which were responsible for the crack propagation. Reasonable results are obtained in terms of the number of loading cycles required to propagate a part-through circumferential crack through the pipe's thickness.

Deposition and characterization of electroless palladium on austenitic stainless steel

A standard electroless plating bath containing PdCl 2 was used to deposit a thin (about 0.5 μm) layer of palladium on type 304 stainless steel specimens. The deposited layer closely followed the contours of the substrate and filled in most surface defects. For typical plating times (10 min), the palladium consisted of a smooth, coherent inner layer of agglomerated nodules and an upper layer of discrete nodules. Removal of the existing oxide layer by sulfuric acid was essential for deposition of palladium. Stirring of the bath led to cracking of the palladium. Earlier work showed that the palladium film was extremely resistant to incorporation of the 60Co radionuclide following exposure in either simulated boiling water reactor or pressurized water reactor primary coolant. Scanning electron micrographs of the palladium coating following exposure in the latter environment showed it to be little changed from the as-deposited film, whereas a duplex layer oxide about 10 μm thick formed on an unplated surface. Thus, the palladium reduced the corrosion of the stainless steel and so reduced the incorporation of 60Co.