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.     

Microstructural analysis and XPS investigation of nodular oxides formed on Zircaloy-4

In order to study nodular oxidation behavior under LOCA conditions, Zircaloy-4 cladding tubes were oxidized in high-temperature steam ambience and the growth evolution and microstructural properties of the oxides formed on the Zircaloy-4 surfaces were then investigated. Optical metallography showed different growth behaviors of uniform black oxide and localized whitish nodular oxide. The changes in composition and chemical states of the oxides formed on the Zircaloy-4 surface were investigated using X-ray photoelectron spectroscopy (XPS). Anion-deficient and non-stoichiometric ZrO was the main species in initial black uniform oxide surface, while stoichiometric ZrO₂ was the main oxide species in the whitish nodular oxide. A stoichiometric composition in the nodular oxide resulted in a decrease in plasticity of the oxide layers. Unlike black uniform oxide, XPS spectrum from the nodular oxide showed clear Sn photopeak, which indicates that Sn species were observed in the nodular oxide only. As a result, it is concluded that the decreased plasticity and localized Sn additives may be the causes of nodular oxide initiation under LOCA conditions.     

Structure of zirconium alloy oxides formed in pure water studied with synchrotron radiation and optical microscopy: relation to corrosion rate

A detailed study was undertaken of oxides formed in 360 °C water on four Zr-based alloys (Zircaloy-4, ZIRLO™,¹ Zr-2.5%Nb and Zr-2.5%Nb-0.5%Cu) in an effort to relate oxide structure to corrosion performance. Micro-beam X-ray diffraction was used along with transmitted light optical microscopy to obtain information about the structure of these oxides as a function of distance from the oxide-metal interface. Optical microscopy revealed a layered oxide structure in which the average layer thickness was inversely proportional to the post-transition corrosion rate. The detailed diffraction studies showed an oxide that contained both tetragonal and monoclinic ZrO₂, with a higher fraction of tetragonal oxide near the oxide-metal interface, in a region roughly corresponding to one oxide layer. Evidence was seen also of a cyclic variation of the tetragonal and monoclinic oxide across the oxide thickness with a period of the layer thickness. The results also indicate that the final grain size of the tetragonal phase is smaller than that of the monoclinic phase and the monoclinic grain size is smaller in Zircaloy-4 and ZIRLO than in the other two alloys. These results are discussed in terms of a model of oxide growth based on the periodic breakdown and reconstitution of a protective layer.     

PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident: Part 2. Formation and destruction of iodine oxides compounds under irradiation - Experimental results modelling

In the case of a severe accident in a nuclear Light Water Reactor (LWR), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core would induce air radiolysis. The air radiolysis products (ARP) could, in turn, oxidise gaseous molecular iodine (I₂) into aerosol-borne iodine-oxygen-nitrogen compounds, abbreviated as iodine oxides (IO_x). These reactions involve the conversion of a gaseous iodine compound resulting in a change of the iodine depletion rate from the containment atmosphere. Kinetic data were produced within the first part of PARIS project on the air radiolysis products formation and destruction. The second part of the PARIS project as presented in this paper deals with the impact of the ARP on the conversion of I₂ into IO_x. The objective was to provide a database to develop new or to validate existing kinetic models of formation and destruction of iodine oxides.The iodine tests of the PARIS project, performed at very low, realistic iodine concentrations, constitute an important database to further develop or validate empirical and mechanistic models on radiolytic I₂ oxidation. In the presence of painted surface areas or silver aerosol surface areas, radiolytic I₂ oxidation is negligible compared to I₂ adsorption on these surfaces for the conditions examined. However, radiolytic I₂ oxidation remains very efficient if surface areas are small or if they are made of the relatively non-reactive stainless steel.     

The Influence of ZrO2 Treatment on the Electrochemical Behavior of Oxygen and Hydrogen on Type 304 Stainless Steels in High Temperature Water

For enhancing the effectiveness of hydrogen water chemistry (HWC) in boiling water reactors (BWRs) in the aspects of lower hydrogen consumption and of a more effective reduction in electrochemical corrosion potential (ECP), the technique of inhibitive protective coating on structural materials was brought into consideration. The application of inhibitive treatment is aimed at deterring the reduction reactions of oxidizing species occurring on metal surfaces and the oxidation reaction of metals. In the current study, electrochemical polarization analyses at 288°C were conducted to characterize the electrochemical properties of ZrO₂ treated and untreated 304 stainless steel specimens in pure water with dissolved oxygen or hydrogen. The polarization results showed that the treated specimens exhibited lower corrosion potentials, corrosion current densities, exchange current densities, and cathodic current densities than the untreated one in high temperature pure water with dissolved oxygen. For the environment with dissolved hydrogen only, reductions in anodic current density and exchange current density were observed, indicating that the ZrO₂ treatment also deterred the oxidation reaction of hydrogen. However, in comparison with the data obtained, the ZrO₂ treatment seemed to be relatively more effective in inhibiting the oxygen reduction reaction than inhibiting the hydrogen oxidation reaction. One additional beneficial outcome was that the anodic current density of the metal was also decreased, leading to a much lower overall corrosion current density of the ZrO₂ treated specimen.     

Leaching test of gamma-emitting Cs,Ru, Zr,and U from neutron-irradiated UO2/ZrO2 solid solutions in non-filtered surface seawater

The leaching behaviors of gamma-ray radionuclides, Cs-137, Ru-103, and Zr-95, produced by neutron irradiation of UO₂/ZrO₂ solid solutions, in real surface seawater were investigated under atmospheric conditions. The fraction of radionuclide inventory leached in the seawater was in the order of Cs > Ru (～U) ? Zr, indicating that the fraction was significantly affected by the chemical state of the radionuclides. However, the amount of soluble nuclides was proportional to that of uranium regardless of whether the solid solutions were prepared under an oxidative or reductive environment. A tiny fraction of Ru was filtered out by a 3 kDa nominal molecular weight cut-off filter after the 160 d leaching test, suggesting a different behavior from its ionic form, but Cs and U did not form a colloid-like species in seawater.     

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%.     

High-temperature oxidation of zircaloy-2 and zircaloy-4 in steam

At temperatures above the $\text{(α + β)}\text{β}$ transformation temperature for zirconium alloys, steam reacts with β-Zr to form a superficial layer of zirconium oxide (ZrO₂) and an intermediate layer of oxygen-stabilized α-Zr. Reaction kinetics and the rate of growth of the combined (ZrO₂ + α-Zr) layer for Zircaloy-2 and Zircaloy-4 oxidation in steam were measured over the temperature range 1050–1850°C. The reaction rates for both alloys were similar, obeyed parabolic kinetics and were not limited by gas phase diffusion. The parabolic rate constants were consistently less than those given by the Baker and Just correlation for zirconium oxidation in steam. A discontinuity was found in the temperature dependence of both the reaction rate and the rate of growth of the combined (ZrO₂ + α-Zr) layer. The discontinuity is attributed to a change in the oxide microstructure at the discontinuity temperature, an observation which is consistent with the zirconium-oxygen phase diagram.     

Prediction of the drying behavior of debris in Fukushima Daiichi Nuclear Power Station for dry storage

Although treatment policies for debris from Fukushima Daiichi Nuclear Power Station have not been decided yet, they may include medium-and long-term debris storage. Dry storage may be desirable in terms of cost and handling, but before implementation, it is necessary to assess hydrogen generation that occurs during storage due to the radiolysis of the water accompanying the debris. Herein, Al₂O₃, SiO₂, ZrO₂, UO₂, and cement paste pellets were prepared as simulated debris with various porosities and pore size distributions. The weight changes of the wet samples were measured at various drying temperatures (100°C, 200°C, 300°C, and 1000°C) via thermogravimetry under helium gas flow (50 cc/min) or reduced pressure conditions (rate: 200 Pa in 30 min), and the resulting drying curves were evaluated. All ceramic pellets exhibited similar drying characteristics in this experiment, indicating that cold ceramics could be used for predicting the drying behavior of ceramic debris. In compariosn with ceramic pellets, cement paste pellets exhibited different behavior and a longer drying time even under 1000°C. In conclusion, it is necessary to decide a standard level for the dry state of molten core concrete interaction (MCCI) products that accompany concrete.     

Leaching behavior of gamma-emitting fission products and Np from neutron-irradiated UO2–ZrO2 solid solutions in non-filtered surface seawater

The gamma ray radionuclides Cs-137, Ba-140, I-131, Ce-141, Ru-103, Zr-95, and Np-239 were produced by neutron irradiation of UO₂–ZrO₂ solid solutions that were synthesized as simulated fuel debris under reducing and oxidizing conditions. The leaching ratio of radionuclides was investigated under atmospheric conditions at 25 °C for non-filtered natural surface seawater, as well as deionized water after filtration with a membrane of 0.45-µm pore size or that of nominal molecular weight limit of 3 kDa. The uranium molar concentration was affected by the oxidation state in the solid solution samples. The congruent dissolution of Cs, I, and Ba with the hexavalent uranium of U₃O₈ was facilitated in the seawater samples, whereas a lower leaching ratio of nuclides was observed in the deionized water samples. Neptunium-239, originally produced from uranium-238 in U₃O₈, showed behavior that was similar to that of Cs, I, and Ba. However, the dissolution of Np (as a parent nuclide of Pu-239) in the debris of UO₂ and UO₂–ZrO₂ was suppressed in the same manner as Zr(IV) and Ce(IV). The concentration exhibited no filtration dependence after 15 d, which shows that most of the leached nuclides can exist in their ionic form in seawater.     

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.     

Hydrogen production in gamma radiolysis of the mixture of mordenite and seawater

Hydrogen production by γ-radiolysis of the mixture of mordenite, a zeolite mineral, and seawater was studied in order to provide basic points of view for the influences of zeolite minerals, of the salts in seawater, and of rise in temperature on the hydrogen production by the radiolysis of water. These influences are required to be considered in the evaluation of the hydrogen production from residual water in the waste zeolite adsorbents generated in Fukushima Dai-ichi Nuclear Power Station. As the influence of the mordenite, an additional production of hydrogen besides the hydrogen production by the radiolysis of water was observed. The additional hydrogen can be interpreted as the hydrogen production induced by the absorbed energy of the mordenite at the yield of 2.3×10^?8 mol/J. The influence of the salts was observed as increase of the hydrogen production. The influence of the salts can be attributed to the reactions of bromide and chloride ions inhibiting the reaction of hydrogen with hydroxyl radical. The influence of the rise in temperature was not significantly observed up to 60°C in the mixture with seawater. The results show that the additional production of hydrogen due to the mordenite had little temperature dependence.     

Zircaloy-4 Cladding Embrittlement due to Inner Surface Oxidation under Simulated Loss-of-Coolant Condition

Embrittlement of Zircaloy-4 cladding by oxidation of the inner surface occurring in an LWR loss-of-coolant accident was studied using simulated fuel containing of A1₂O₃ pellets sheathed in Zircaloy-4 specimen cladding, filled with Ar gas, and sealed. This simulated fuel rod was heated from outside until the isothermal oxidation temperature between 880 and 1,167°C was obtained after the cladding burst. This exposed the inner surface of the cladding to the environmental atmosphere, provided by steam flowing at a constant rate in the range of 0.13–1.6 g/cm²-min.

The embrittlement of the specimen due to inner surface oxidation is influenced primarily by the amount of hydrogen absorbed by the Zircaloy-4. Ring compression tests conducted at 100°C on test pieces constituted of sliced sections of oxidized specimen showed that Zircaloy containing more than 200–300 wt.ppm of absorbed hydrogen became brittle when oxidized at temperatures above 1,000°C. In the range of oxidation temperature 932 to 972°C, brittleness did not appear below 500–750 wt.ppm absorbed hydrogen.

Hydrogen absorbed by the Zircaloy precipitated in the form of fine hydride crystals formed along previous β-phase grain boundaries. Peaks were found in the distribution of hydrogen absorbed on the inner surface, at a distance of 15–45 mm upward and downward of the rupture opening. Within this range, the distance was influenced by the oxidation temperature and steam flow rate.     

Comparison of the air oxidation behaviors of Zircaloy-4 implanted with yttrium and cerium ions at 500 °C

As a valuable process for surface modification of materials, ion implantation is eminent to improve mechanical properties, electrochemical corrosion resistance and oxidation behaviors of varieties of materials. To investigate and compare the oxidation behaviors of Zircaloy-4, implantation of yttrium ion and cerium ion were respectively employed by using an MEVVA source at the energy of 40 keV with a dose ranging from 1×10¹⁶ to 1×10¹⁷ ions/cm². Subsequently, weight gain curves of the different specimens including as-received Zircaloy-4 and Zircaloy-4 specimens implanted with the different ions were measured after oxidation in air at 500 °C for 100 min. It was obviously found that a significant improvement was achieved in the oxidation behaviors of implanted Zircaloy-4 compared with that of the as-received Zircaloy-4, and the oxidation behavior of cerium-implanted Zircaloy-4 was somewhat better than that of yttrium-implanted specimen. To obtain the valence and the composition of the oxides in the scale, X-ray photoemission spectroscopy was used in the present study. Glancing angle X-ray diffraction, employed to analyze the phase transformation in the oxide films, showed that the addition of yttrium transformed the phase from monoclinic zirconia to tetragonal zirconia, yet the addition of cerium transformed the phase from monoclinic zirconia to hexagonal zirconia. In the end, the mechanism of the improvement of the oxidation behavior was discussed.     

Leaching of actinide elements from simulated fuel debris into seawater

For the prediction of the leaching behavior of actinide elements contained in the fuel debris that has arisen from the severe accident in Fukushima Daiichi Nuclear Power Station (NPS), a simulated fuel debris consisting of UO₂–ZrO₂ solid solution doped with ¹³⁷Cs, ²³⁷Np, ²³⁶Pu and ²⁴¹Am tracers was synthesized, and agitated leaching tests were conducted for the simulated fuel debris in seawater. The synthesized simulated fuel debris was immersed and shaken in natural seawater collected at a coast 11 km away from Fukushima Daiichi NPS. The brief leaching test conditions were T = 25 °C and solid–liquid ratio = 4 g/l, and the test duration was up to 31 days. The ratio of tracers leached into seawater from the simulated fuel debris by the agitated leaching test for 4 days was evaluated to be 0.09% for U, 0.01% for Np, 0.01% for Pu, 0.01% for Am and 35.39% for Cs by the α or γ spectrometry of the soluble fraction. The leaching of actinides from the real fuel debris in reactor units 1–3 in Fukushima Daiichi NPS is expected to be suppressed in comparison with that from normal light water reactor spent fuel.     

Pore structure in oxide films on irradiated and unirradiated zirconium alloys

Scanning electron microscopy, two-stage replica electron microscopy, and porosity measurements on oxide films formed on Zircaloy-2 and Zr-2.5% Nb alloy samples are analyzed to show the differences in the pore structures in the two cases. For Zircaloy-2 the evidence indicates that the pore network that develops in the post-transition oxidation regime penetrates to the oxide-metal interface, whereas for the Zr-2.5% Nb alloy the evidence suggests that an impervious barrier layer of oxide persists to very high weight gains, although the outer part of the oxide is porous. No distinct visual differences were seen between oxide films formed in the laboratory and in-reactor. The differences in the pore structures of the oxides formed on Zircaloy-2 and Zr-2.5% Nb may account for the large differences in in-reactor behaviour, particularly hydrogen absorption, seen with pressure tubes of the two alloys.     

Oxidation kinetics and related phenomena of zircaloy-4 fuel cladding exposed to high temperature steam and hydrogen-steam mixtures under PWR accident conditions

With respect to the behavior of Nuclear Pressurized Water Reactor fuel cladding during accidents the oxidation kinetics of Zircaloy-4 tubing in steam and hydrogen-steam mixtures and the related changes in the mechanical properties have been investigated. Short tube sections were exposed to steam in surplus between 600 and 1600°C under isothermal and temperature transient conditions and to steam of limited supply and hydrogen-steam mixtures between 800 and 1300°C under isothermal conditions. Tube capsules were creep-rupture tested in steam under isothermal/isobaric (600–1300°C, 7–150 bar) and various temperature/pressure transient conditions.The oxidation kinetics of Zircaloy-4 is described in case of short-term reaction by simple rate laws. The long-term behavior is proved to be highly influenced by the breakaway transition and finally by total consumption of the tube wall. On the basis of the isothermal behavior the short-term temperature transient case can be understood and also modelled. Deviations from predictable behavior are the consequences of Zr- and ZrO₂-phase transformations. The protective character of preexistent scales is lost above 1100°C for the same reason.The limits of steam supply rate dependent oxidation and its interrelation with the competing hydrogen take-up are described.The consequences of oxidation on the mechanical properties are increased strength and decreased ductility, whereas the response to creep deformation is an overall increase in oxidation due to oxide cracking.     

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

The effects of hydrazine on the corrosion of Zircaloy-2 were examined in supercritical water. Hydrazine could be used as a reducing agent to control the corrosive environment for the coolant of boiling water reactors (BWRs). Before the corrosion test, the applicability of supercritical water for corrosion testing of zirconium alloys was studied. Supercritical water was found to be a useful solvent for testing corrosion based on the following facts: (1) the weight gain of Zircaloy-2 in supercritical water followed the same cubic law with the activation energy of 133 kJ/mol as that in water and steam did, and (2) the weight gain in supercritical water at 723 K and 24.5 MPa was more than 8 times greater than that in water at 561 K and 7.8 MPa depending on immersion time. The corrosion tests in supercritical water at 723 K and 24.5 MPa under γ-irradiation for 1,000 h were conducted to study the effects of adding nitrogen and ammonia on the corrosion of Zircaloy-2. Nitrogen and ammonia are decomposed products of hydrazine. The measured weight gain, oxide film thickness, and amount of hydrogen pick-up had slight differences between cases with and without the additives. Based on these data, it was concluded adding hydrazine to the coolant has little influence on the corrosion of Zircaloy-2 used in BWR cores.     

Particulars of radiolysis and mass transfer in two-phase water coolant of a nuclear power facility with steam generation at an interphase surface

A model of an interphase transfer of stable products of the radiolysis of water in boiling coolant is developed taking account of the intensity of their delivery to the interphase boundary in the liquid phase and removal into the vapor phase with vapor generation on the interphase surface. A computational study is made of the radiolysis of the coolant and interphase transfer of the products of the radiolysis of water in the core and on the pulling section of BWR of the Oskarshamn-2 nuclear power plant in Sweden. A comparison of the computational data with the results of the technical measurements of the coolant composition of the BWR at the Oskarshamn-2 nuclear power plant showed that the accumulation of stable products of the radiolysis of water in the vapor-gas phase of the coolant is determined by the kinetics of radiolysis in the liquid phase, the concentration of the oxygen-containing components in the liquid phase is due to the present of hydrogen peroxide in it.     

Developments in PWR cladding corrosion: materials and models

In pressurized water reactors Zircaloy-4 is a standard fuel cladding material. The aim of this paper is to present and evaluate corrosion data generated both in-reactor, and out-of-reactor on PWR claddings made of both Zircaloy-2 and Zircaloy-4 materials. The oxide thickness measurements of cladding tubes irradiated in the Ringhals 3 reactor, and oxide weight gain measurements carried out in Sandvik autoclaves at 400°C, 10.3 MPa clearly show that the stress relief annealed Zircaloy-2 is more corrosion resistant than Zircaloy-4 produced with an identical fabrication route. Furthermore, autoclave tests indicate that the hydrogen pickup fraction of the two alloys is very similar. The obtained data have been evaluated in regard to chemical composition and heat treatment. In addition, computer models, which simulate thermal and hydraulic reactor conditions and corrosion kinetic processes simultaneously, have been used to predict the in-reactor corrosion behaviour of the claddings.