Effects of seawater components on radiolysis of water at elevated temperature and subsequent integrity of fuel materials

Effects of seawater components on radiolysis of water at elevated temperature have been studied with a radiolysis model and a corrosion test under gamma-ray irradiation conditions to evaluate the subsequent influence on integrity of fuel materials used in an advanced boiling water reactor. In 2011, seawater flowed into the nuclear power plant system of the Hamaoka Nuclear Power Station Reactor No. 5 during the plant shutdown operation. The reactor water temperature was 250 °C and its maximum Cl^? concentration was ca. 450 ppm when seawater was mixed with reactor water. The radiolysis model predicted that the main radiolytic species were hydrogen, oxygen and hydrogen peroxide. Concentrations of radiolytic products originating from Cl^? and other seawater components were found to be rather low. The dominant product among them was ClO₃^? and its concentration was found to be below 0.01 ppm for a 10⁵ s irradiation period. No significant corrosion of zircaloy-2 and 316L stainless steel was found in the corrosion test. These results led to the conclusion that the harmful influence of radiolytic products originating from seawater components on integrity of fuel materials must be smaller than that of Cl^? which is the main ionic species in seawater.     

Effects of gamma-ray irradiation on crevice corrosion repassivation potential of stainless steel in high temperature diluted simulated seawater

The crevice corrosion repassivation potentials (E_R,CREV) of type 304 stainless steel (304 SS) were measured in high temperature (373–553 K), diluted simulated seawater under gamma-ray irradiation, in order to confirm the effects of gamma-ray irradiation on the crevice corrosion behavior of a representative stainless steel in seawater. Overall, for high temperatures, the E_R,CREV values decreased with increasing chloride ion concentration, which was the same as the behavior observed under the non-irradiated condition. The E_R,CREV values measured under gamma-ray irradiation were the same or slightly higher than E_R,CREV values measured under the non-irradiated condition when the [Cl^?] was the same. Consequently, it was confirmed that the threshold potential of crevice corrosion of 304 SS for the gamma-ray irradiation of 1.8 kGy at least did not deteriorate compared with the non-irradiated condition. Under the conditions of this work (seawater composition, [Cl^?] range, dose rate, absorbed dose, flow rate, etc.), the crevice corrosion of 304 SS could be suppressed by maintaining the potential below the threshold potential which was determined approximately as ?0.3 V vs. SHE even for the irradiated condition at temperatures up to 553 K.     

Effect of gamma-ray irradiation on the deoxygenation of salt-containing water using hydrazine

In spent fuel pools at the Fukushima Daiichi nuclear power plant, hydrazine was added to salt-containing water in order to reduce dissolved oxygen. Hydrazine is known to reduce dissolved oxygen in high-temperature pure water, but its deoxygenation behavior in salt-containing water at ambient temperature in the presence of radiation is unknown. Deoxygenation using hydrazine in salt-containing water was thus investigated using a ⁶⁰Co gamma-ray source and artificial seawater at room temperature. Water samples containing a small amount of hydrazine were irradiated at dose rates of 100–10,000 Gy/h. The concentration of dissolved oxygen in the water samples was measured before and after irradiation. Notably, a decrease in the dissolved oxygen was only observed after irradiation, and the dissolved oxygen concentration decreased with increasing dose rate and irradiation time. The rate of decrease in the amount of dissolved oxygen using hydrazine was slow in the presence of salts. Kinetic considerations suggested that the deoxygenation of the salt-containing water exposed to gamma-ray irradiation using hydrazine was suppressed by chloride ions.     

Characterization of the stress corrosion cracking behavior of thermally sensitized 20Cr-25Ni stainless steel in a simulated cooling pond environment

Niobium stabilized 20Cr-25Ni stainless steel is used for nuclear fuel cladding in the UK's fleet of advanced gas cooled reactors (AGRs). The cladding can have chromium-depleted grain boundaries as a consequence of irradiation in a reactor core, rendering a small proportion of cladding susceptible to intergranular stress corrosion cracking in cooling pond waters after removal from the reactor. In this work, thermal sensitization was used to simulate chromium depletion and the sensitized material was assessed for its susceptibility to pitting corrosion and stress corrosion cracking using slow strain rate testing (SSRT). Elevated chloride concentrations were used to accelerate corrosion initiation and propagation. In 10 ppm chloride and 80 °C, the pitting potential was at potentials between +375 mV and +400 mV (SCE). SSRT appeared to lower the pitting potential, with intergranular corrosion and intergranular stress corrosion cracks observed to nucleate at potentials of +200 mV (SCE).     

Influence of oxide structure on the elevated temperature pitting behaviour of Fe-Cr-Ni alloys

The influence of temperature, potential, alloy and electrolyte composition on the growth of hot water oxide layers on stainless steel type AISI 321 and Fe-Cr-model alloys is presented. The correlation between the oxide layer properties and the pitting corrosion behaviour is discussed. In situ electrochemical investigations have been carried out in aqueous electrolytes (pH 8) on hydrothermal conditions varying from 150 to 250°C. After hot water exposition the steel surfaces were examined microscopically. The porosity of these layers increases with rising temperature due to a change in crystallinity from a fine-crystalline to a coarse-crystalline structure. It is shown that the pitting initiation mechanism is strongly dependent on the morphology of the oxide layers. At 150°C pitting corrosion occurs, whereas at 250°C the formation of extended hollow shaped corrosion has been observed. The cause of the detected inhibiting effect of sulphate ions on the chloride induced pitting is discussed. Furthermore, evident dependencies of the pitting susceptibility on the chromium content of the alloy and on the oxide layer preformation potential have been observed.     

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, H₂O₂ and H₂, 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.     

候选材料在650 ℃超临界水中的腐蚀行为研究

研究了3种候选材料（347、HR3C和In-718）在650 ℃、25 MPa去离子水中的均匀腐蚀行为,使用场发射扫描电镜（FEG-SEM）和能谱（EDS）观察了不同腐蚀时间的表面氧化膜形貌与合金元素分布,使用掠入射X射线衍射（GIXRD）分析了氧化膜相结构。结果表明,3种材料腐蚀失重均符合抛物线规律,347的失重为HR3C和In 718的40倍以上;3种材料氧化膜均以Ni(Cr, Fe) ₂O₄为主,In-718点蚀严重,347氧化膜明显脱落,HR3C氧化膜较均匀致密;高温超临界水中,提高合金的Cr含量有助于增强均匀腐蚀性能,添加Nb有损合金的点蚀抗力。     

Effect of seawater components on corrosion behavior of SUS316L in nitric acid solution containing metal ions

During the accident that occurred at the Fukushima Daiichi nuclear power plant, a large volume of seawater was introduced as coolant into the storage pools for spent nuclear fuel. If this fuel is reprocessed, some components of seawater will be mixed with the nitric acid solution containing metal ions in the reprocessing process where stainless steels are used as structural material. In this study, we investigated the effect of seawater components in high active liquid waste (HAW) containing nitric acid and metal ions as fission products on the corrosion behavior of SUS316L stainless steel.

Corrosion tests were conducted in surrogate HAW containing artificial seawater (ASW). Intergranular corrosion was observed in the HAW with ASW, where Ru increased the corrosion potential to the transpassive region. An increase in the amount of ASW led to a decrease in the corrosion rate and suppression of intergranular corrosion. Interactions between Ru ions and seawater components, such as chloride ions, were indicated by the results of extended X-ray absorption fine structure spectroscopy and cyclic voltammetry analyses of the solution containing ASW and HAW.     

Simulation for radiolytic products of seawater: effects of seawater constituents,dilution rate,and dose rate

Radiolysis calculations of simulated seawater were conducted using reported data on chemical yields and chemical reaction sets to predict the effects of seawater constituents on water radiolysis. Hydrogen, oxygen, and hydrogen peroxide were continuously produced from simulated seawater during γ-ray irradiation. The concentration of H₂ exceeded its saturation concentration before it reached the steady-state concentration. The production behavior of these molecules was significantly promoted by the addition of bromide ions (Br^?) because of the high reactivity of Br^? with the hydroxyl radical, an effective hydrogen scavenger. It is also shown that the concentrations of these molecules were effectively suppressed by diluting seawater constituents by less than 1%.     

Effect of radiolytic hydrogen peroxide in tritiated water on the first passivity of titanium

Titanium is an interesting metallic material for nuclear applications. This is due to its passivity ensured by a compact and chemically stable oxide film that spontaneously covers the metal surface. This work aims at studying the electrochemical behavior of titanium in the presence of tritiated water at pH 4 and containing hydrogen peroxide generated by radiolytic reaction. Tritium in tritiated water always causes difficulties in corrosion. The corrosion potential can be either in the active or prepassive regions depending on the concentration of radiolytic oxidizing species and intermediate species formed on the surface in the active region. Therefore, the behavior of titanium was studied by cyclic voltammetry and electrochemical noise in situ to provide an indication of mechanisms, transient formation and instabilities of oxide in active region. In the prepassivity and passivity, according to the results of electrochemical impedance spectroscopy and coulometric curves, titanium is protected by a semiconductor layer essentially formed of TiO₂. Between the region of first and second passivity, oxide consists of a dielectric layer and a semiconductor layer. These two layers protect the titanium against corrosion.     

Preselection of Ni-Cr(-Mo) alloys as potential canister materials for vitrified high active nuclear waste by electrochemical testing

Several Ni-Cr(-Mo) alloys (Hastelloy C4, Inconel 625, Sanicro 28, Incoloy 825, Inconel 690) were tested by electrochemical methods to characterize their corrosion behavior in chloride containing solutions at various temperatures and pH-values in respect to their application as canister materials for final radioactive waste storage. Especially, Hastelloy C4 was tested by potentiodynamic, potentiostatic and galvanostic measurements. As electrolytes H₂SO₄ solutions were used, as parameters temperature, chloride content and pH-value were varied.All tested alloys showed a clearly limited resistance against pitting corrosion phenomena; under severe conditions even crevice corrosion phenomena were observed. The best corrosion behavior, however, is shown by Hastelloy C4, which has the lowest passivation current density of all tested alloys and the largest potential region with protection against local corrosion phenomena.     

Behaviour of spent HTR fuel elements in aquatic phases of repository host rock formations

One back-end option for spent HTR fuel elements proposed for future HTR fuel cycles in the EC is an open fuel cycle with direct disposal of conditioned or non-conditioned fuel elements. This option has already been chosen in Germany due to the political decision to terminate the use of HTR technology. First integral leaching investigations at Research Centre Juelich on the behaviour of spent HTR fuel in salt brines, typical of accident scenarios in a repository in salt, proved that the main part of the radionuclide inventory cannot be mobilised as long as the coated particles do not fail. However, such experiments will not lead to a useful model for performance assessment calculations, because a failure of the coatings by corrosion will not occur during experimental times of a few years. In order to get a robust and realistic model for the long-term behaviour in aqueous phases of host rock systems, it is necessary to understand the barrier function of the different parts of an HTR fuel element, i.e. the matrix graphite, the different coating materials, and the fuel kernel.Therefore, our attention is focused on understanding and modelling the barrier performance of the different parts of an HTR fuel element with respect to their barrier function, and on the development of an overall model for performance assessment. In order to understand this behaviour, it is necessary to start with investigations of unirradiated material, and to proceed with experiments with external gamma irradiation to determine the effects of oxidising radiolysis species. Further experiments with irradiated material have to be performed to investigate the influence of the irradiation damage, and finally an investigation has to be made of the irradiated material plus additional gamma irradiation. Experimental data are now available for the diffusive transport of radionuclides in the water-saturated graphite pore system, the corrosion rates of unirradiated graphite with and without external gamma irradiation and unirradiated and irradiated silicon carbide, and for the dissolution rates of UO₂ and (Th,U)O₂ fuel kernels with and without external gamma irradiation. All investigations were performed in aquatic phases from salt, granite, and clay host rock.     

Effects of hydrazine addition and N2 atmosphere on the corrosion of reactor vessel steels in diluted seawater under gamma-rays irradiation

Seawater was injected into the reactor cores in the Fukushima Daiichi Nuclear Power Station. Corrosion of primary containment vessel (PCV) steel and reactor pressure vessel (RPV) steel is considered to progress until the molten fuel debris is removed. To evaluate durability of the PCV and RPV steels, corrosion tests were conducted in diluted seawater at 50 °C under gamma-rays irradiation of dose rates of 4.4 and 0.2 kGy/h. To evaluate the effect of hydrazine (N₂H₄) as an oxygen scavenger under gamma-rays irradiation, 10 and 100 mg/L N₂H₄ were added to the diluted seawater. Without addition of N₂H₄, weight loss in the PCV and RPV steels irradiated with the 0.2 kGy/h dose rate was comparable with those without irradiation and weight loss in the vessel steels irradiated with the 4.4 kGy/h dose rate was higher than those without irradiation. Under irradiation, weight loss in the PCV and RPV steels in diluted seawater containing N₂H₄ was comparable with that in diluted seawater without N₂H₄. When gas phase in the flask was replaced with N₂, weight loss in the PCV and RPV steels, and O₂ and H₂O₂ concentrations in the diluted seawater decreased.     

Effects of Hydrogen Peroxide on Intergranular Stress Corrosion Cracking of Stainless Steel in High Temperature Water, (V)

The difference in electrochemical corrosion potential of stainless steel exposed to high temperature pure water containing hydrogen peroxide (H₂O₂) and oxygen (O₂)is caused by differences in chemical form of oxide films. In order to identify differences in oxide film structures on stainless steel after exposure to H₂O₂ and O₂ environments, characteristics of the oxide films have been examined by multilateral surface analyses, e.g., X-ray diffraction (XRD), Rutherford back scattering spectroscopy (RBS), secondary ion mass spectroscopy (SIMS) and X-ray photoelectron spectroscopy (XPS). Preliminary characterization results of oxide films confirmed that the oxide film formed under the H2O2 environment consists mainly of hematite (α-Fe₂O₂), while that under the O₂ environment consists of magnetite (Fe₃O₄). Furthermore oxidation at the very surface of the film is much more enhanced under the H₂O₂ environment than that under the O₂ environment. It was speculated that metal hydroxide plays an important role in oxidation of stainless steel in the presence of H₂O₂. The difference in electric resistance of oxide film causes the difference in anodic polarization properties. It is recommended that several anodic polarization curves for specimens with differently oxidized films should be prepared to calculate ECP based on the Evans diagram.     

Distributions of density and fission products in the reaction product between irradiated MOX fuel and molten zircaloy-2

Two- and three-dimensional images were obtained by X-ray CT in the reaction product between zircaloy-2 cladding tube and MOX fuel. The gamma-ray intensity distributions in the same specimen were also obtained by gamma-ray measurements of two fission products (Cs-137 and Eu-154) and one neutron-activated nuclide (Co-60). The average values of the fuel density (about 10.5 g/cm³) and the cladding density (about 6.55 g/cm³) were obtained in the metallic phase region by evaluation of the density distributions on two-dimensional X-ray CT images. The distributions of the crushed fuel pellet and the pores were also clearly observed in the three-dimensional X-ray CT images. The following results were found from the gamma-ray measurement. First, Cs-137 was observed in the unreacted fuel region and the pore region in the metallic phase region. Second, Eu-154 was widely distributed to all regions. Finally, Co-60 was confirmed only in the metallic phase region.     

Chloride induced stress corrosion cracking of candidate canister materials for dry storage of spent fuel

Susceptibility to chloride induced stress corrosion cracking (ESCC) of candidate canister materials, UNS S31260 and UNS S31254 stainless steels (SS), was investigated by a constant load test in air at temperatures of 343 and 353 K with relative humidity (RH) of 35%, and at 373 K without controlling RH. UNS S31260 and UNS S31254 SS did not fail until 37,700 h at 353 K with RH = 35%, where UNS S30403 SS failed within 250–500 h. The same tendency also was obtained at 343 K, suggesting the superior ESCC resistance of UNS S31260 and UNS S31254 SS. Even rust was not observed on the specimens tested at the temperature of 373 K. To explain the higher ESCC resistance, the pitting potential was measured in the saturated synthetic sea water at temperatures from 303 to 353 K, since ESCC is usually associated with localized corrosion such as pitting and may be closely related to the corrosion resistance. The pitting potentials of UNS S31260 and UNS S31254 SS were much higher than that of UNS S30403 SS. Thus, it was concluded that the superior ESCC resistance is attributable to the higher resistance of UNS S31260 and UNS S31254 SS to pitting corrosion. The critical relative humidity for ESCC, under which no ESCC occurs, is equal to or higher than 15% at temperatures < 353 K judging from ESCC behavior of UNS S30400 SS.     

辐照对堆用锆合金腐蚀行为的影响

综述了堆用锆合金辐照腐蚀研究的进展，包括概况、机理和模型，在ＰＷＲ运行环境下，辐照确实增强了锆合金的腐蚀，其增强因子大于２－３倍。为了降低运行成本而提高卸料燃耗和提高运行温度，使得辐照腐蚀加剧；从不同侧面提出了辐照增强腐蚀的机理和模型，包括不同射线与材料的相互作用机理，氧化膜中缺陷的形成，组织结构的变化，电导的变化和脆化，金属的辐照损伤，中间相的辐照分解，水的辐照分解和辐照对反应活化能的影响等；近年来关于中间相在不同射线下辐照分解的研究较为活跃，其进展结果为从机理上建立辐照损伤参量与辐照腐蚀性能参量的关系提供了依据。     

Intergranular stress corrosion cracking susceptibility of sensitized type 304 steel in 561 K pure water under γ-ray irradiation

Intergranular stress corrosion cracking (IGSCC) of sensitized type 304 stainless steel has been investigated in 561 K water under γ-ray irradiation at a flux of 2.6 × 10³ C kg⁻¹ h⁻¹ by slow-strain-rate tensile tests. The IGSCC susceptibility was enhanced by γ-ray irradiation in water containing 8 ppm dissolved oxygen (DO). The DO dependence of the IGSCC susceptibility was observed in the water under γ-irradiation. Although slight IGSCC susceptibility was observed even in deaerated water (less than 1 ppb DO) under γ-ray irradiation, the susceptibility was completely suppressed by injection of hydrogen into the water. The enhancement of IGSCC susceptibility seems to be related to the formation of H₂O₂ in high temperature water by radiolysis under γ-ray irradiation and the H₂O₂ formation rate is markedly decreased by hydrogen injection.     

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.     

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).