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.     

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%.     

Hydrogen generation by water radiolysis with immersion of oxidation products of Zircaloy-4

In order to predict the hydrogen gas generation from seawater or water in which debris would be included by the severe accident of nuclear power plant, we investigated the effect of ZrO₂ and the oxidation products of Zircaloy-4 on hydrogen gas generation by radiolysis of water since the radiolytic generation could be affected by materials immersed in water. Powders of well-characterized oxides and oxidation products were immersed in either seawater or distilled water, and irradiated by gamma ray from a Co-60 source. The observed hydrogen yield, G(H₂), was measured as a function of the weight fraction of oxide in water up to 50 wt%. The enhancement of the hydrogen generation by radiolysis of water with the commercial oxides and the oxidation products of Zircaloy-4 was quite small or absent in seawater. But the enhancement was observed in the presence of the oxides or the oxidation products at low weight fraction in distilled water. This enhancement in distilled water seemed to be dependent on specific surface area or particle size, but its dependence on the crystal structure was not apparent in the experimental results. The enhancement was saturated at higher ZrO₂ weight fractions and it was not apparent in the seawater.     

Characterization and storage of radioactive zeolite waste

For the safe storage of zeolite wastes generated by the treatment of radioactive saline water at the Fukushima Daiichi Nuclear Power Station, this study investigated the fundamental properties of herschelite adsorbent and evaluated its adsorption vessel for hydrogen production and corrosion. The hydrogen produced by the herschelite sample is oxidized by radicals as it diffuses to the water surface and thus depends on the sample's water level and dissolved species. The hydrogen production rate of herschelite submerged in seawater or pure water may be evaluated by accounting for the water depth. From the obtained fundamental properties, the hydrogen concentration of a reference vessel (decay heat = 504 W) with or without residual pure water was evaluated by thermal–hydraulic analysis. The maximum hydrogen concentration was below the lower explosive limit (4%). The steady-state corrosion potential of a stainless steel 316L increased with the absorbed dose rate, but the increase was repressed in the presence of herschelite. The temperature and absorbed dose at the bottom of the 504 W vessel were determined as 60 °C and 750 Gy/h, respectively. Under these conditions, localized corrosion of a herschelite-contacted 316L vessel would not immediately occur at Cl^? concentrations of 20,000 ppm.     

Verification of hydrogen isotope separation by pressure swing adsorption process: Successive volume reduction of isotopic gas mixture using SZ-5A column

For the purpose of verifying the applicability of pressure swing adsorption (PSA) process to such as volume reduction of tritiated waste storage, an experimental series was carried out by a PSA apparatus having a zeolite packed column operated at the liquefied nitrogen temperature, where synthetic zeolite 5A was used as a candidate of adsorbents. Experimental results are shown here which were obtained from cyclic operation of isolating a volume of hydrogen decontaminated with its heaver isotope from a mixture of H₂ and D₂ while reducing a volume of this mixture storage. Successive reduction during six cycles is observed in the inventory of this hydrogen mixture in a gas holder. Experimental data are analyzed in order to evaluate the performance of this PSA process operating the hydrogen isotope separation, where several factors are introduced defining efficiencies of decontamination, volumetric reduction, and so on. These factors suggest that the PSA process is available for successive reduction of a tritiated hydrogen storage inventory. A tritium waste management system of PSA process combined with electrolysis is considerable which is aiming at reducing the inventory of tritiated water in storage.     

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.     

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.     

Effect of gamma radiolysis on pit initiation of zircaloy-2 in water containing sea salt

In spent fuel pools at the Fukushima Daiichi Nuclear Power Station (1F), seawater was injected for cooling purposes after the tsunami disaster in March 2011. It is well known that the chloride in the seawater has the potential to cause localized corrosion (e.g., pitting corrosion) in metals. In this study, we evaluated the pitting potentials of zircaloy-2, the material used in the fuel cladding tubes in 1F, as a function of chloride concentration. To accomplish this, we used artificial seawater under gamma-ray irradiation and investigated the effect of radiolysis on pit initiation of zircaloy-2 in water containing sea salt. Changes in the composition of water containing sea salt were analyzed as well, both before and after gamma-ray irradiation. The characteristics of the resultant oxide films formed on zircaloy-2 were evaluated by X-ray photoelectron spectroscopy and electrochemical impedance spectroscopy. The experimental results showed that the pitting potential under irradiation was slightly higher than that under conditions in which no radiation was present, and that the pitting potential decreased with increasing chloride concentration in the presence as well as the absence of radiation. Solution analyses for water containing sea salt showed that hydrogen peroxide was generated by irradiation. The oxide film was composed of zirconium oxide and was made thicker during the irradiation. The higher pitting potential could thus be explained by the capacity of hydrogen peroxide to oxidize the surface and enhance oxide film formation. Under gamma-ray irradiation, the zircaloy-2 surface with an oxide film formed by radiolysis products was found to be resistant to pitting in the presence of chloride.     

Catalytic mitigation of hydrogen risk during wet transportation of radioactive materials

Abstract

Major challenges in the area of wet transportation of radioactive materials are reliability and safety of transportation casks. In most cases, the bottom part of the cask is filled with water whereas a gaseous mixture is contained in the upper part. In such a configuration, water radiolysis leads to the formation of hydrogen and oxygen, which continuously enriches the gaseous mixture. Among the functions to be satisfied, wet transportation systems shall thus allow the control of the hydrogen content below its flammability limit. This is currently achieved by limiting the transportation duration so as to reopen the cask before the critical hydrogen concentration is reached. Development of new technologies that would mitigate the hydrogen risk is all the more motivated because it would allow an extension of the transportation duration. AREVA-TN International and the Institut de recherches sur la catalyse et l'environnement de Lyon have developed a catalytic system which aims at buffering the hydrogen concentration far below the flammability limit. The principle of this catalyst is to recombine the hydrogen with the oxygen formed by water radiolysis. The present paper gives an overview of the development of this catalytic recombining system. It describes the laboratory qualification tests undertaken for the evaluation of the recombining efficiency. Particular attention is placed on the recombining efficiency after immersion of the catalyst in borated water, which would occur in a nuclear reactor pool during loading of used fuel. Laboratory investigations, carried out in an autoclave simulating a transportation cask, showed that, after immersion in borated water, the catalytic system allows the recombination of 3% hydrogen in less than 24 h at temperatures as low as 35°C.     

“Protoka” Experimental Stand for Studying High-Temperature Radiolysis of Water by Accelerated Protons

The experimental stand Protoka, put into operation in 2000 at the Russian Science Center Kurchatov Institute, for studying high-temperature radiolysis of water under the action of accelerated protons is described. The principles of simulation of radiolytic effects under the action of nuclear reactor radiation on water, using the results of experiments on radiolysis of water by accelerated protons, are discussed. The Protoka data on the radiolysis of water solutions of hydrazine and a mixture of molecular hydrogen and hydrogen peroxide are presented.     

Trapping of radiolytic hydrogen by amorphous cobalt oxysulfide

Hydrogen production from the radiolysis of liquid and gaseous hydrocarbons was studied in the presence of several transition metal sulfides. Cobalt oxysulfide obtained by aqueous precipitation was the most efficient admixture to decrease radiolytic production of hydrogen by pure hydrocarbons or mixtures of saturated and unsaturated hydrocarbons. Cobalt oxysulfide was characterized by XRD, scanning and transmission microscopy, and IR spectroscopy. It seems to be amorphous compound with the impurities of lamellar Co(OH)₂ phase. The organic phases were analysed before and after irradiation by γ-rays or protons in order to elucidate the origin of the effect. It has been shown that the solid does not change the composition neither the amount of the organic radiolysis products, neither in liquid nor in gas phase experiments. Therefore the presence of solid does not influence any radiolysis processes in the organics. Amorphous cobalt oxysulfide acts essentially as a trap of hydrogen, being able to absorb considerable amounts of H₂ (up to 0.5 mol H₂/at. Co). The study of the solid-gas interaction showed that slow reaction of cobalt oxysulfide with hydrogen occurs at ambient conditions independently of the irradiation of the system.     

超临界水的辐化反应研究

It is known that water exists in supercritical state above 374C and 22.1MPa, where gas phase and liquid phase are merged into a single phase. In the supercritical state, the density is controllable by changing the pressure. The properties such as the ionic product, solubilities of salts, gas and organic compounds, and dielectric constant of supercritical water are very different from those of water at room temperature.Recently much attention has been paid to supercritical water because many possible applications such as synthesis of functional materials, waste oxidation, and biomass conversion have been proposed by using above peculiar properties of the supercritical water. Thus, much intensive work is in progress all over the world. In addition, a new concept of nuclear reactor using the supercritical water as a coolant has been proposed, which has been chosen by DOE, USA as one of the forth generation nuclear reactors.Therefore, radiation chemistry study of the supercritical water seems inevitably important, yet little work has been done so far. We started radiation chemistry study of supercritieal water by pulse radiolysis and γ radiolysis five years ago. We have observed hydrated electrons as a first target. We also measured inorganic radicals, metal ions, organic radicals in high temperature and supercritical water by a pulse radiolysis technique. It was found that absorption spectra of the transient species are dependent on temperature. Some transients show red-shift, and some bands are blue-shifted, with increasing temperatures. A G-value evaluation of water decomposition products was also done by using methyl viologen as a scavenger and it was revealed that the values are significantly dependent not only on temperature but also on density in supercritical water. An extended study on the behavior of solvated eleclions in different alcohols at high temperatures and super critical state is also in progress.     

A study of low-energy ion induced radiolysis of thiol-containing amino acid cysteine in the solid and aqueous solution states

The radiolysis of cysteine under plasma discharge and irradiation of low-energy ion beam was investigated. The damage of cysteine in aqueous solution under discharge was assessed via the acid ninhydrin reagent and the yield of cystine produced from the reaction was analyzed by FTIR. In addition, the generation of hydrogen sulfide was also identified. The destruction of solid cysteine under low-energy ion beam irradiation was estimated via monitoring IR bands of different functional groups (–SH, –NH₃, –COO^?) of cysteine, and the production of cystine from ion-irradiated solid cysteine after dissolution in water was also verified. These results may help us to understand the inactivation of sulphydryl enzymes under direct and indirect interaction with the low-energy ion irradiation.     

Adsorption capacity of hydrogen isotopes on mordenite

In a fusion reactor system, a monitoring of hydrogen isotopes including tritium is necessary for the safety of system control and operation. A gas chromatography using a cryogenic separation column is one of the methods for hydrogen isotope analysis. Synthesis zeolite such as molecular sieve 5A (CaA) is a candidate material of the separation column, and its property varies by the ratio of silica to alumina, the kinds of cation and so on. If the factor affected the hydrogen adsorption property of the synthesis zeolite is clarified, it may lead to the development of the new zeolite optimized to the separation column. So, in this work, adsorption capacity of hydrogen (H₂) and deuterium (D₂) for mordenite (MOR) and NaY type zeolite (NaY) were investigated at various temperatures, and were compared with CaA. The amount of adsorption per unit weight of MOR was larger than that of CaA, and that of NaY was smaller than that of CaA. The adsorption isotherms were expressed by sum of two Langmuir equations, and the Langmuir coefficients of H₂ and D₂ were proposed. Furthermore, the Langmuir coefficients of HD, HT, DT and T₂ were estimated by the reduced mass. The correlation between the adsorption properties and the physical parameters of the zeolite were not confirmed.     

Molecular-sieving effect of zeolite 3A on adsorption of H2, HD and D2

Synthetic zeolite 3A has the molecular-sieving windows of nominal diameter 0.3 nm in its crystal lattice framework, which obstruct the crystalline adsorption of molecules of diameter larger than 0.3 nm, except water, hydrogen and helium. The window's diameter slightly varies with temperature, however, that is endorsed in experimental results that hydrogen cannot be adsorbed at the liquid-nitrogen temperature. Authors measured the range of temperature where zeolite 3A permits hydrogen adsorption, and revealed the temperature difference of several degrees in appearance of molecular sieving for H₂ and D₂. This difference is important because from a H₂–D₂ mixture one isotope could be isolated by adsorption if operated at a temperature regulated between the molecular-sieving appearance temperatures. We have reported large values of D₂/H₂ separation factor obtained from molecular-sieving experiments. In this study, the effect of sieving for the hybrid-atomic isotope HD is examined using a H₂–HD–D₂ mixture. We here report the experimental HD/H₂ separation factor evaluated between the D₂/H₂ factor and unity. This result is significant because where the effective molecular diameter concerning the sieving mechanism is suggested. From this knowledge, the isotopic effect of sieving for HT and DT can be predicted.     

Computerised Modelling of Radiolysis and Corrosion Effects in Transport Containers for Fuel Spent Nuclear

Abstract

The radiolysis of water and/or gases within transport containers for spent nuclear fuel may result in the generation of hydrogen and oxygen gases and also the enhanced corrosion of the materials in contact with the water. These effects are important, particularly when the fuel container is also used for storage post-transport prior to reprocessing or disposal. The behaviour of a range of radiolytic systems has been studied. Plant behaviour has been simulated in numerous laboratory experiments: plant and experimental results have been linked by a computerised model describing the radiolysis mechanism and predicting the quantities and production rate of gaseous and corrosive species. This allows prediction of plant performance over a long time scale. The model is based on a well-accepted radiolysis mechanism supplemented with specific measurements made at the Harwell laboratory. Model capabilities include inert atmospheres, materials corrosion, variations in water and gas volumes or aqueous chemistry. The model has been applied to design stage radiolysis assessments of transport containers; information from operating plant has been interpreted to advise on design improvement, e.g. diminution of gas production using easily corroded scavengers to remove oxygen. Radiolysis in gas filled dry storage containers for spent nuclear fuel has been studied; corrosive product production (e.g. nitric acid), which is important for fuel cladding integrity has been assessed. The development and use of this computerised model is described with a current summary.     

PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident: Part 1. Formation and destruction of air radiolysis products—Experimental results and modelling

In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models.The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating (“paint”) and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments.It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these “non-noticeable influence” could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations.The PARIS results are then analysed using two different kinetic models, an empirical and a mechanistic one. The kinetic constants within an empirical model including formation and destruction of air radiolysis products, derived from PARIS results, are in reasonable agreement with those determined previously for lower steam fractions.From the mechanistic model IODAIR-IRSN, it is concluded that ozone is the predominant air radiolysis product at low doses in air/steam atmospheres. At doses higher than 1 kGy, nitrogen dioxide becomes increasingly important, both due to an increase in its concentration and due to a simultaneous decrease in ozone concentration.     

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.     

Adsorption Behavior of Americium on Zeolites

Zeolites are expected to be used in buffer materials as an admixture to bentonite to increase their ion exchange ability. The behavior of ion exchange adsorption of Am³⁺ on various zeolites was examined by batch and column methods. The distribution coefficient of Am³⁺, K_d, increased with equilibrium pH from 1.5 to 4.0. L zeolite and mordenite having large pore size yielded relatively large values of K _d over 10³ at pH 3.0. The K _d for L zeolite was unaffected by the concentration of coexisting cations, i.e. Na⁺ up to 0.1 M, and K⁺, Ca²⁺ and Mg²⁺ less than 0.01 M. The Am³⁺ ion was not detected in the effluent passed through the L zeolite column up to 90 bed volumes, even when the feed solution contained a relatively large amount of Na⁺-0.5 M. Americium ion adsorbed on zeolites could be quantitatively eluted with 0.1 M nitric acid.     

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

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