Radiolytic Gases Research Articles

A three-dimensional criticality safety analysis code for spent fuel solution system

The nuclear criticality safety analysis is crucial for the nuclear safety of spent fuel reprocessing plants. Thus, a set of numerical tools with high accuracy and efficiency to predict the criticality and simulate the hypothetical accidents in the reprocessing procedure is of great importance. Among them, this paper focus on the analysis of the storage tank which is used for dissolution and storage of the spent fuel. According to the characteristics of spent fuel solution system, a parallel 3D critical safety analysis tool for fissile solution system, Hydra-TD, is developed. Generally, it contains the cross-section generation model, the three-dimensional space-time neutron kinetics model, and the R-Z two-dimensional heat conduction and radiolysis gas simulation model. The verifications based on the experiments of the SILENE and TRACY facility are conducted. The results show good accuracy of the prediction of the first fission power peak, multiplication time and total fission energy, indicating the reliability of these numerical models.

Рецензия на горно-геологическую часть обоснования строительства подземного хранилища радиоактивных отходов

The city district “Closed Administrative-territorial formation of the city of Zheleznogorsk” has appointed and organized public discussions on the materials of the justification of the license (including environmental impact assessment materials) for activities in the field of nuclear energy use “Creation of a second-class RW storage facility at FSUE GCK”. The mining and geological materials on the construction site of an underground repository of radioactive waste (RW) of the second hazard class in Zheleznogorsk of the Krasnoyarsk Territory are considered. The purpose of the construction of the storage facility is to ensure the storage of second-class radioactive waste at the site of the mining and chemical plant in volumes that meet the needs of the enterprise until the national operator is ready to handle radioactive waste. The review is a response to an offer to discuss supposedly public materials of open access. The subject of public discussions was conducted in the form of a survey. It is proposed to carry out the initial stage of the justification of the object on the basis of the Law “On Subsoil”. The following conclusion is made: the target horizon after the conservation of the object is likely to be flooded, primarily along vertical collector trunks; heterogeneous backfilling of the internal free space of trunks and horizontal workings with finished concrete lining with crushed/granular natural bentonite will not help; in reality, the volume of water in the burial zone will significantly determine the characteristics and the inexhaustible resource of the overlying water supply layers of rocks, rather than rocks of the target range, and even without recharge from other sources, the flooded PGZRO will quickly turn into a supplier of hot water and radiolytic gases. The opinion is expressed that it is unacceptable to turn a temporary storage facility into a permanent disposal site for radioactive waste.

Radiolysis of bituminized radioactive waste: a comprehensive review

In the realm of radioactive waste management, the impact of radiolysis on bitumen and bituminized radioactive waste also called bituminized waste product (BWP) is considered one of the most significant factors influencing structural changes and the generation of radiolysis gas bubbles. This review provides a comprehensive overview of several studies that have explored the intricate interaction between radiation and various types of bitumen, to gain a better understanding of how such waste ages when exposed to radiation. While these studies provide insights into the diverse effects of radiolysis on bitumen, they also highlight numerous unanswered questions. The absence of gas bubbles does not necessarily indicate an absence of gas production, leaving further research to be undertaken. The complexities of bitumen radiolysis offer multiple avenues for future investigation, aiming to enhance our understanding and provide comprehensive solutions for bituminous radioactive waste management. Among the studied types of bitumen, blown R85/40 bitumen and straight-run distilled bitumen with a penetration grade of 70/100 stands out due to their widespread use in immobilizing co-precipitated radioactive sludges. These bituminous matrices play an essential role in understanding the broader implications of radiolysis within the context of bituminous waste management. This review underscores the significance of further research into radiolysis and bitumen ageing, emphasizing the need for a more in-depth exploration of these complex phenomena and their implications for the long-term safety and efficacy of repositories and disposal facilities.

STORAGE: A Source Term Model for Intermediate-Level Radioactive Waste

The French Alternative Energies and Atomic Energy Commission and Orano have developed a modeling tool named the Simulation TOol Of RAdiolysis Gas Emission (STORAGE) for assessing gas generation of intermediate-level waste (ILW). The first version of this model was designed to estimate gas (more specifically hydrogen) production by radiolysis of organic and water-containing materials. The code deals with different types of waste packages: metal drums, concrete drums, bitumen packages, and compacted waste containers. Diverse radioactive waste can be handled: spent fuel cladding, reprocessing sludge, contaminated technological waste (gloves, bags, bottles, etc.), ion exchange resin, etc. The validity of the model was evaluated using a series of measurements performed on U,Pu–contaminated solid waste from the Orano plutonium laboratories at the MELOX facility. A benchmark study for compacted waste containers was also implemented; the results of STORAGE were compared with reference calculations performed by Orano Projets. Future improvements of the STORAGE model are also presented.

Adsorption Behavior of Pb, Cd, Ca, Al, Zn And Co Metal Ions on Cation Exchange Resin Dowex 50 Wx8 (Nh4+ Form) From Aqueous Acetone-Ammonium Propionate Media

There has a wide interest in stability of different types of ion exchange resins against ionizing radiations. The irradiation damage and stability of polystyrendigni benzene and phenolformaldehide sulphonic acid ion exchange resins was investigated by determination of the radiolytic gases. The gas analysis showed a very good linear dose dependence of the evolved hydrogen and carbon monoxide. The linearity of carbon dioxide was good for Dowex 50 WX 10 but less pronounce for zeo-carb 215. Sulphur dioxide was at least partially produced by direct action and was consumed by secondary reaction during irradiation. Kuzin et al [1] analysed radiolytic gases evolved from a carboxylic acid exchange resin. Wiely et al [2] investigated gamma radiation induce degradation of sulphonetic styrle resin cross link with 4 and 8 % m / p dinyl benzene isomer or mixtures thereof including commercial divenyle benzene have been shown loss in capacity 5-12. 8% at the radiation intensities of 290 R/hr and total dose 0.91 – 1.0 x 10-8 rad in presence of water.

Hydrogen Generation from α Radiolysis of Organic Materials in Transuranic Waste. Comparison Between Experimental Data and STORAGE Calculations

The Alternative Energies and Atomic Energy Commission and Orano have developed a modeling tool named the Simulation Tool Of RAdiolysis Gas Emission (STORAGE) for assessing gas generation of intermediate-level waste. The first version of this model was designed to estimate gas (more specifically hydrogen) generation by radiolysis of organic materials contained in waste packages. To verify the validity of the model, a series of measurements was performed on U, Pu–contaminated solid waste issued from the Orano plutonium laboratories at the MELOX facility. Twenty-one drums containing technological waste (gloves, bags, filters, metallic parts, etc.) packaged inside polyvinyl chloride sleeves were set up and hydrogen production was measured over a period of more than 1 year. Several levels of contamination and organic content were studied. STORAGE calculations are conservative and most of the time in good agreement with experimental measurements with the uncertainties. As expected, the simplest cases (organic waste or filtering media) are well described by the model. The data are obviously more widely dispersed when the waste is composed of a mixture of organic materials and metal. Nevertheless, an understanding of the waste (package composition) allows a fairly precise description and ultimately a satisfactory estimation of the hydrogen production rate.

Performance tests of catalysts for the safe conversion of hydrogen inside the nuclear waste containers in Fukushima Daiichi

The safe decommissioning as well as decontamination of the radioactive waste resulting from the nuclear accident in Fukushima Daiichi represents a huge task for the next decade. At present, research and development on long-term safe storage containers has become an urgent task with international cooperation in Japan. One challenge is the generation of hydrogen and oxygen in significant amounts by means of radiolysis inside the containers, as the nuclear waste contains a large portion of sea water. The generation of radiolysis gases may lead to a significant pressure build-up inside the containers and to the formation of flammable gases with the risk of ignition and the loss of integrity.In the framework of the project “R&D on technology for reducing concentration of flammable gases generated in long-term waste storage containers” funded by the Japanese Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT), the potential application of catalytic recombiner devices inside the storage containers is investigated. In this context, a suitable catalyst based on the so-called intelligent automotive catalyst for use in a recombiner is under consideration. The catalyst is originally developed and mass-produced for automotive exhaust gas purification, and is characterized by having a self-healing function of precious metals (Pd, Pt and Rh) dissolved as a solid solution in the perovskite type oxides. The basic features of this catalyst have been tested in an experimental program. The test series in the REKO-4 facility has revealed the basic characteristics of the catalyst required for designing the recombiner system.

Behaviour of magnesium phosphate cement-based materials under gamma and alpha irradiation

Stabilization and solidification of low- and intermediate-level radioactive waste using Portland cement, possibly blended with fly ash or blastfurnace slag, is a well-established practice. However, when the waste contains high amounts of alpha emitters, this solution can be restricted by the strong release of radiolytic gases, wherein H2 is the most abundant. This work investigates the interest of using magnesium potassium phosphate cement (MPC), a binder with a high chemical water demand, as a possible substitute to Portland cement (PC). The radiolytic gas production of PC and MPC pastes and mortars is determined under external gamma and internal alpha irradiation. The H2 radiolytic yield of MPC materials is found to be 2 to 3 times smaller than that of PC references, provided that the main part of the mixing water is consumed by K-struvite formation. Moreover, gamma irradiation of a MPC mortar up to an integrated dose of 10 MGy has no significant influence on its mechanical strength (flexural, compressive) nor on its mineralogy. MPC materials are thus potential candidates for the conditioning of high amounts of radioactivity with limited H2 release. The H2 production of MPC materials can be reduced further by adding radical scavengers or H2 getters within the matrix. However, other radiolytic gases such as O2 are often produced, making these solutions potentially less attractive considering the concern of pressure build-up within the cemented waste package.

A Cellular Ceramic as a Catalyst Support for Radiolytic Gas Recombination Devices

Catalytic recombination of hydrogen and oxygen as radiolysis gases was studied using ceramic blocks of cellular structure as a catalyst support. The characteristics of the catalyst samples obtained were compared to those of the presently used granulated catalyst RK-102-12 (OPK-2). The possibility of preventing the transition of the catalytic recombination of hydrogen and oxygen to the flame combustion mode by making a catalyst with “gradient” activity was examined. The results obtained confirm good prospects for using the cellular ceramic as a catalyst support for passive hydrogen recombiners to ensure hydrogen safety at nuclear power plants.

Reactivity Feedback Effect on Supercritical Transient Analysis of Fuel Debris

Transient analysis for possible prompt supercritical accidents of fuel debris in the Fukushima Daiichi Nuclear Power Station is quite important. However, unlike solution fuel systems, there is little knowledge about supercritical transient analysis in fuel debris systems. In particular, reactivity feedback effects, which may have a significant impact on the results of the analysis, are important and require further study. In particular, the impacts of radiolysis gas void and moderator boiling should be discussed. Thus, the purpose of this study is to clarify whether the reactivity feedback effects of radiolysis gas and boiling of the moderator impact the supercritical transient analysis in fuel debris systems. To accomplish this, we used a power profile obtained by the MIK code with the Doppler reactivity feedback effect; radiolysis gas analysis and heat transfer analysis were performed. For the radiolysis gas analysis, the AGNES2 model was modified to consider the difference between solution fuel and fuel debris systems. The heat transfer analysis used an OpenFOAM solver to perform conjugate heat transfer calculations. We found that the radiolysis gas void was negligible when probable G values, which are the generation number of molecules per absorbed energy, were used. In addition, the results showed that boiling could be also negligible under most conditions. However, we found that the boiling time may be earlier than the peak time of the power when the radius of the fuel debris particle is small. In this case, ignoring the boiling may give conservative results. These considerations should be included in future analyses.

Numerical simulation of the long-term radiolytic gas accumulation in diversified measurement pipeline for BWR using GASFLOW-MPI

Unforeseen radiolytic gas combustions have occurred in many Boiling Water Reactors (BWRs) and caused severe damages in the past. This combustion is caused by the long-term accumulation of radiolytic combustible gas in a pipeline with a top dead-end due to the main steam continuous condensation. The main objective of this study is to conduct a numerical simulation for a diversified measurement pipeline for BWR with GASFLOW-MPI, in order to evaluate its the long-term radiolytic gas accumulation. Firstly, the model for the radiolytic gas absorption in the condensing film is developed in GASFLOW-MPI and is validated with the experiment of “Hamaoka pipe”. The calculated temperature profiles along the pipe centerline agree well with the experimental data. Then a 3D model for part of the diversified measurement pipeline of BWR is built and analyzed with GASFLOW-MPI to evaluate the long-term radiolytic gas accumulation. Moreover, two factors are discussed in this paper, one is the local condensation enhancement at the floating chambers which is, different from the rest part of the pipeline, not covered by insulator, and the other is the radiolytic gases absorption model. The result shows that there will not be strongly local accumulation of radiolytic gases at the “dead-end” of the pipeline, because the momentum of the reservoir can be balanced with the buoyancy caused by the density difference. Moreover, a mass balance can be established when the mass of radiolytic gases sucked from the reservoir is balanced with that absorbed in the film when considering the absorption model. As a result, the concentration of the radiolytic gases accumulated is far below the combustible limit for the pipeline.

Influence of formulation parameters of cement based materials towards gas production under gamma irradiation

The release of radiolysis gas is a concern that may restrict the use of cement materials to condition intermediate level radioactive waste. Indeed, water naturally present in cement materials produces hydrogen gas (which can be explosive/flammable under some conditions) when it is exposed to ionizing radiation.The primary goal of the MATRICE (MAterials Resistant to Irradiation based on Cement) project is to identify and define formulations of cement materials in order to minimize the quantities of hydrogen gas released by radiolysis. The first approach is the minimization of water amount in standard Portland materials (calcium silicate-based cements) by addition of specific compounds (superplasticizers) to enable the preparation of wasteform. The second approach is to use “alternative” cement such as calcium sulfoaluminate cement. This cement was expected to release less hydrogen because the quantity of water needed for cement hydration is higher than Portland and moreover, their hydrates differ from those of hydrated calcium silicate mostly encountered in Portland based materials.Based on gamma irradiations with a60Co source, the results obtained demonstrate that the first approach is efficient but yet limited because the production of hydrogen of Portland pastes is about proportional to the total amount of water present in the materials. Thus, a tremendous drop of hydrogen production cannot be reach because rheological constraint does not allow a huge reduction of water, even with efficient superplasticizers. The second approach using calcium sulfoaluminate cements as an alternative binder provides results that are quite similar to Portland cement concerning the production of hydrogen under gamma irradiation.

ESS Target Water Cooling, Purification and Radiolysis Gas Handling

The goal for the ESS facility is to construct an accelerator with the ability to deliver a 5 MW proton beam on the tungsten target. Most of the power is deposited as heat in the fluids and structures in and around the target. To be able to run the facility the nuclear heating from the spallation process need to be cooled in a safe and reliable way. The tungsten target in itself is helium cooled, but this system is in turn water cooled and there are several other water cooling systems, e.g. for moderators, reflectors and structural components. In this paper the water cooling systems are described and what challenges and requirements the systems needs to deal with.

Synthesis, Characterization and Application of Thermostable Hydrophobic Pt Catalysts for Oxidation of H2

The studies were aimed at the development of thermostable hydrophobic Pt catalyst for hydrogen oxidation in the close-to-stoichiometric mixture with oxygen to provide the high gas conversion at the temperature not higher than 353 K in the direct contact with water. The catalyst is to be used for utilization of radiolytic gases (hydrogen and oxygen) at atomic power stations. The studies were focused on the Pt/Al 2 O 3 type catalyst with a modified surface. Modified γ-Al 2 O 3 was used for preparation of catalysts containing 0.5 wt% of platinum. The catalyst hydrophobicity, thermal stability, texture, surface composition, catalytic performance for hydrogen oxidation, and operational stability were studied during 50-hour testing. The established properties of the catalyst were thermal stability up to 773 K, superhydrophobic surface, reaction rate constants no less than 4±1 s –1 (at 333 K). The oxidation efficiency was no less than 99.999 % at the hydrogen flow rate of 50 L/h. The results obtained demonstrate potentialities of the prepared hydrophobic catalyst and of the technology for hydrogen oxidation over this catalyst.

γ-Ray radiolysis of acetohydroxamic acid in HNO3 and its radiolytic product

Acetohydroxamic acid (AHA) is a novel salt-free reagent used for the separation of Pu and Np from U in the advanced Purex process. This paper reports the γ-ray damage of AHA in HNO3 and its radiolytic product. For 0.2 mol L−1 AHA in 0.2–2.0 mol L−1 HNO3 irradiated at a dose of 5–25 kGy, the radiolytic rate of AHA is 6.63–77.5%, and it increases with the HNO3 concentration and absorbed dose. The main radiolytic gases are N2O and H2, with volume fractions of (0.500–16.2) × 10−2 and (1.30–11.8) × 10−3, respectively, and they increase with the absorbed dose; the H2 volume fraction decreases with increasing HNO3 concentration. The main liquid radiolytic products are CH3COOH and HNO2, and their concentrations are (3.40–19.7) × 10−2 and (0.200–4.80) × 10−3 mol L−1, respectively, which increase with the HNO3 concentration. Since a significant concentration of HNO2 is present in the irradiated AHA-HNO3 solution, a holding reductant must be used to destroy HNO2 and stabilize Pu(III) and Np(V) when AHA is applied for the separation of Pu and Np from U.

The characterization of radioactive waste: a critical review of techniques implemented or under development at CEA, France

This review paper describes the destructive and non-destructive measurements implemented or under development at CEA, in view to perform the most complete radioactive waste characterization. First, high-energy photon imaging (radiography, tomography) brings essential information on the waste packages, such as density, position and shape of the waste inside the container and in the possible binder, quality of coating and blocking matrices, presence of internal shields or structures, presence of cracks, voids, or other defects in the container or in the matrix, liquids or other forbidden materials, etc. Radiological assessment is then performed using a series of non-destructive techniques such as gamma-ray spectroscopy, which allows characterizing a wide range of radioactive and nuclear materials, passive neutron coincidence counting and active neutron interrogation with the differential die-away technique, or active photon interrogation with high-energy photons (photofission), to measure nuclear materials. Prompt gamma neutron activation analysis (PGNAA) can also be employed to detect toxic chemicals or elements which can greatly influence the above measurements, such as neutron moderators or absorbers. Digital auto-radiography can also be used to detect alpha and beta contaminated waste. These non-destructive assessments can be completed by gas measurements, to quantify the radioactive and radiolysis gas releases, and by destructive examinations such as coring homogeneous waste packages or cutting the heterogeneous ones, in view to perform visual examination and a series of physical, chemical, and radiochemical analyses on samples. These last allow for instance to check the mechanical and containment properties of the package envelop, or of the waste binder, to measure toxic chemicals, to assess the activity of long-lived radionuclides or pure beta emitters, to determine the isotopic composition of nuclear materials, etc.

Radiolytic Dissolution of Calcite under Gamma and Helium Ion Irradiation

Samples of calcite have been irradiated using both the 60.7 MeV helium ion beam of the ARRONAX (Nantes, France) cyclotron and 661.7 keV gamma radiations. The irradiation of calcite pellets results mainly in radiolytically generated gas (H2, CO2). The presence of superficial water during the calcite irradiation in a humid atmosphere is a key factor for the production of radiolysis gases. Besides, the irradiation of calcite–water biphasic media was performed. The gas released after these irradiations is quantified. No significant amount of CO2 has been detected, whereas carbonate ions act as OH• scavengers, thus inducing an increase of H2 production. The amount of hydrogen peroxide created confirms this scavenging mechanism. Finally, ionic chromatography experiments with irradiated solutions allow us to quantify the organic anions (formate HCOO–, acetate CH3COO–, propionate CH3-CH2COO–, and oxalate C2O42–) formed by calcite and/or carbonate ion radiolysis and study the variation of calcium carbonate solubil...

Gamma radiolysis of anion exchange resins based on 4-vinylpyridine in aqueous solution

Radiation-induced degradation of the weakly and strongly 4-vinylpyridine basic ion exchange resins by gamma radiolysis was investigated in the presence of air and liquid water. This study is focused on evaluating the radiolytic gases (H2, CO, CO2 and CH4) and liquid products (water-solute TOC and NH4 +). The weakly basic resin yielded lower amounts of H2 and CO and higher amounts of CO2 than those of the strongly basic resin. Moreover, the strong basic resin tended to yield greater amounts of NH4 +. Resins were characterized by the FTIR spectroscopy technique and the results showed that the resins structures are relatively stable.

Gamma radiation effect on gas production in anion exchange resins

Radiation-induced decomposition of Amberlite IRA400 anion exchange resin in hydroxide form by gamma radiolysis has been studied at various doses in different atmospheres (anaerobic, anaerobic with liquid water, and aerobic). The effect of these parameters on the degradation of ion exchange resins is rarely investigated in the literature. We focused on the radiolysis gases produced by resin degradation. When the resin was irradiated under anaerobic conditions with liquid water, the liquid phase over the resin was also analyzed to identify any possible water-soluble products released by degradation of the resin. The main products released are trimethylamine (TMA), molecular hydrogen (H2g) and carbon dioxide (CO2g). TMA and H2g are produced in all the irradiation atmospheres. However, TMA was in gaseous form under anaerobic and aerobic conditions and in aqueous form in presence of liquid water. In the latter conditions, TMAaq was associated with aqueous dimethylamine (DMAaq), monomethylamine (MMAaq) and ammonia (NH4+aq). CO2g is formed in the presence of oxygen due to oxidation of organic compounds present in the system, in particular the degradation products such as TMAg.

Effect of γ irradiation on nuclear bituminized waste products (BWP): X-ray microtomography and rheological characterization

In the last 50 years bitumen has been widely used in France as a matrix for embedding low- and intermediate-level radioactive waste. This organic matrix is subjected to α, β and γ radiation generating radiolysis gases, which consist mainly of hydrogen but also some carbon oxide and lighter hydrocarbons. The gases form bubbles whose growth entails swelling of the bitumen if the radiolytic process of gas generation is more important than the removal process. To assess the influence of γ radiation on the swelling and rheological behavior of the bitumen matrix, external γ irradiation experiments were performed on pure bitumen (PB) and on a model synthesized bituminized waste product (BWP: 60 wt% bitumen + 40 wt% salts). Greater swelling was observed for the BWP samples than for PB. X-ray microtomography shows that the hydrogen bubbles are more numerous in the case of BWP, indicating that the salts may be considered as preferential nucleation sites. Moreover, the rheological characterization of the samples shows that the Newtonian viscosity is higher for the BWP model sample, and that the irradiation enhances this phenomenon. In the case of accelerated irradiation (dose rate exceeding the actual dose rate), swelling increases with the irradiation dose since the gas production rate is higher than the removal process until a plateau is reached at approximately 4–5 MGy, indicating an equilibrium between gas production and release. Finally, the impact of the dose rate on the number, size and shape of bubbles and on macroscopic swelling is discussed on the basis of X-ray microtomography results.