Radiolytic Gas Production Research Articles

Radiolytic Production of Fluorine Gas from MSR Relevant Fluoride Salts

Radiolytic fluorine gas production at temperatures of 40°C to 60°C was investigated for the fluoride salts LiF, BeF2, UF4, ThF4, and 71.7LiF-16BeF2-12.3UF4 (FliBe-UF4) by gamma irradiation of powdered samples using spent fuel elements from the High Flux Reactor (HFR) Petten as the irradiation source; work of a similar nature was previously performed at Oak Ridge National Laboratory in the period 1965 to 1995. Gamma irradiation was conducted for just over 41 days, with total absorbed gamma dose ranging from ~45 MGy for the lightest salts to ~170 MGy for ThF4 and UF4. By measuring the gas pressure within salt-filled capsules during irradiation, it was possible to quantify radiolytic gas production for all salt samples except UF4. Production rates are reported as the salt G-values, measured as number of fluorine molecules produced per 100 eV of energy absorbed (molecules F2/100 eV). The G-values of the salts were found to be G(LiF) ~0.004, G(BeF2) ~0.009, G(ThF4) ~0.021, and G(FLiBe-UF4) ~0.005.

Wetting-induced volumetric collapse of UO2 powder beds and the consequence on transient nuclear criticality excursions

Mathematical and computational models are proposed to simulate wetting-induced volumetric collapse of fissile powder beds. Slumping, nuclear thermal hydraulics, radiolytic gas, and steam production models are coupled with point neutron kinetics to investigate transient nuclear criticality excursions in two 5-wt% enriched UO2 fissile powder beds with varying levels of wetting-induced volumetric collapse. The two beds are distinguished by their mean powder particle size of 30 μm and 100 μm. For the UO2 powder beds modelled, the re-distribution of UO2 powder and moderator due to slumping introduced a negative reactivity into the system. This increased the amount of time taken for a delayed critical state to be reached once infiltration began, and also reduced the total fission energy generated over the course of the simulated transient. The total fission energy generated ranged from 42 MJ to 48 MJ 100 seconds after the initial nuclear criticality excursion was observed for the 30μm sized UO2 powder bed. The fission energy of the larger sized powder bed (100 μm), varied from 42 MJ to 57 MJ. Larger discrepancies between the slumped and un-slumped initial peak power are predicted. Peak powers varied from 29.2 MW to 106 MW for the smaller-sized powder particles, whereas for larger particles, the peak powers varied from 255 MW to 501 MW.

Radiolytic Gas Production from Aluminum Coupons (Alloy 1100 and 6061) in Helium Environments-Assessing the Extended Storage of Aluminum Clad Spent Nuclear Fuel.

Corrosion of aluminium alloy clad nuclear fuel, during reactor operation and under subsequent wet storage conditions, promotes the formation of aluminium hydroxide and oxyhydroxide layers. These hydrated mineral phases and the chemisorbed and physisorbed waters on their surfaces are susceptible to radiation-induced processes that yield molecular hydrogen gas (H2), which has the potential to complicate the long-term storage and disposal of aluminium clad nuclear fuel through flammable and explosive gas mixture formation, alloy embrittlement, and pressurization. Here, we present a systematic study of the radiolytic formation of H2 from aluminium alloy 1100 (AA1100) and 6061 (AA6061) coupons in “dry” (~0% relative humidity) and “wet” (50% relative humidity) helium environments. Cobalt-60 gamma irradiation of both aluminium alloy types promoted the formation of H2, which increased linearly up to ~2 MGy, and afforded G-values of 1.1 ± 0.1 and 2.9 ± 0.1 for “dry” and “wet” AA1100, and 2.7 ± 0.1 and 1.7 ± 0.1 for “dry” and “wet” AA6061. The negative correlation of H2 production with relative humidity for AA6061 is in stark contrast to AA1100 and is attributed to differences in the extent of corrosion and varying amounts of adsorbed water in the two alloys, as characterized using optical profilometry, scanning electron microscopy, Raman spectroscopy, and X-ray diffraction techniques.

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.

An extension of the point kinetics model of MIPR to include the effects of pressure and a varying surface height

This work aims to expand upon a previous point kinetics model of an aqueous homogeneous reactor and make several additions to it including the effects of average pressure, plenum gas composition and temperature, vertical advection of solution due to voidage and thermal expansion and a neutron flux profile which varies in both time and height from the reactor base. It is found that these additions cause a small change to the reactor behaviour but the main advantage is to allow the modelling of different scenarios including failure of the gas management system which normally removes hydrogen and fission product gases from the plenum gas. Such a failure is seen to lead to a rapid increase in hydrogen content with implications for reactor safety. The model is also used to simulate the CRAC-43 experiment and, with the addition of simple models for the delay of radiolytic gas production due to the need to saturate the fuel solution and sloshing of the fuel solution, is found to approximate the experimental results well.

A Two-Dimensional Multiregion Computer Model for Predicting Nuclear Excursions in Aqueous Homogeneous Solution Assemblies

The reprocessing of nuclear fuel usually involves the process of chemical separation. The fuel, usually in oxide form, is first dissolved in some type of acid such as nitric, sulfuric, or hydrofluoric. This results in the fuel being transformed into a homogeneous aqueous fissile solution. In this form there may be a higher probability of an accidental criticality of the solution, especially when being transported through pipes or stored in vessels. Here, a new two-dimensional computer model for simulating power and pressure pulses in aqueous fissile solutions has been developed. This model includes a radiolytic gas production model that tracks the number of gas bubbles produced during an excursion. An equation of state has been developed that accounts for the production of inertial pressure due to a lag in thermal expansion and the creation of radiolytic gas bubbles. In addition, a study of various reactivity feedback mechanisms occurring during nuclear bursts has been made. The model`s predicted power and pressure pulses are compared with data from the KEWB and SILENE solution pulsed reactor experiments and have produced results that closely match the experimental data and that exhibit the main features of the experimental power and pressure traces.

Comparative behaviour under external gamma irradiation of ion exchanger waste solidified into epoxide or epoxide cement matrices

In France, IER solidification processes, by embedding into polymer based matrices, are operated at the industrial scale. The CEA process utilizes the epoxide resin as IER embedding matrix, with an embedding ratio equal to 0.5. A new embedding process was studied, using an epoxide cement compound matrix, with an embedding ratio 0.4. On this new material and IER embedded forms, an external gamma irradiation test was carried out, using a Cobalt 60 irradiator, called IRCA, until absorption of a 3 MGy integrated gamma dose. The effects of the gamma radiolysis were measured and the results compared to those obtained on IER epoxide embedded forms. The radiolytic gas production, dimensional changes and other secondary effects are discussed.

Compared Study of Radiolysis-Induced Gas Liberation in Rocksalt from Various Origins

ABSTRACTThe evolution of impure rocksalt samples under irradiation has been investigated and compared to the evolution of pure halite, concerning the radiolytic release of gases. Four different types of rocksalt (pure halite, anhydrite-bearing, sylvinite-bearing, and marneous) have been irradiated with spent fuel at doses ranging from 104Gy to 107Gy, under synthetic air or helium at 50°C, 150°C, and 200°C. Marneous and anhydrite-bearing salt produced higher amounts of gas than the two other types, especially for H2, CO, and CH4. Radiolytic gas production essentially originates from the decomposition of traces of different kind of impurities : fluid inclusions, which can be composed of brine and/or organic fluids and gases; minerals such as carbonates, sulfates, or hydrated minerals; organic matter, in general (kerogen). Pure halite decomposes only at very high doses (more than 107Gy) giving off some Cl2 and other corrosive gases. Gas production increases steadily with dose for H2, CO, and CH4, contrary to what is observed for the pure Asse salt, where H2 and CO, respectively, peak at 106, and 105 Gy. The increase, with dose, in the production of CH4, C2H6, and C3H8 is far higher for marneous salt and higher for anhydrite-bearing salt. The decrease of the O2/N2 (oxygen consumption) ratio in the irradiation atmosphere is in the following order: marneous>anhydrite>halite=sylvinite. The overall effect of γ-irradiation on impure rocksalt can be described as follows : destruction of organic matter that produces chemically reduced gases (H2, CO, hydrocarbons), after an initial stage of oxygen consumption, which is reached more or less quickly, depending on the availability and quantity of organic matter. The use of impure, i.e, organic matter-bearing salt, as host-rock or engineered barrier for nuclear wastes, may be questioned, because of the high amount of gases produced in the immediate vicinity of the container (explosion hazards, pressure build-up). The present study will be followed by a modelling of the in-situ radiolytic gas production, taking into account: 1) the dose distribution with time, 2) the temperature distribution with time, and 3) the geometry of the repository, and will allow us to estimate the potentiel safety impact of radiolytic gas production in the case of a repository emplaced in impure rocksalt.

Radiolysis experiments for the 'NET ASCB': preliminary results and further plans at ‘SCK/CEN’ Mol

Neutron irradiations of 4.35 M LiNO 3and 2.175 M Li 2SO 4 solutions under nearly ASCB relevant conditions resulted in the following observed radiation chemical yields, expressed as the number of molecules or ions that are formed per 100 eV absorbed energy: for the LiNO 3 solution, G(H 2) = 0.22 (±0.03), G(O 2) = 1.7 (±0.2) and G(NO 2 −) = 2.4 (±0.3); for the Li 2SO 4 solution, G(H 2) = 1.6 (±0.1) and G(O 2) = 0.18 (±0.08). The 90% confidence limits are given between brackets. These G-values and the linear relationship between radiolytic gas production and absorbed energy still have to be confirmed at higher pressures and at higher dose rates. Apart from further closed capsule irradiations, an in-pile loop is planned at SCK/CEN Mol for an integrated study of radiolysis, chemistry control and corrosion aspects of the NET ASCB.

Radiolytic gas formation in high-level liquid waste solutions

High-level fission product waste solutions originating from the first-cycle raffinate stream of spent fast breeder reactor fuel reprocessing have been investigated gas chromatographically for their radiolytic and chemical gas production. The solutions showed considerable formation of hydrogen, carbon dioxide and dinitrogen oxide, whereas atmospheric oxygen was consumed completely within a short time. In particular, carbon dioxide resulted from the radiolytic degradation of entrained organic solvent. After nearly complete degradation of the organic solvent, the influence of hydrazine and nitrogen dioxide on hydrogen formation was investigated. Hydrazinium hydroxide led to the formation of dinitrogen oxide and nitrogen. After 60 d, the concentration of dinitrogen oxide had reduced to zero, whereas the amount of nitrogen formed had reached a maximum. This may be explained by simultaneous chemical and radiolytic reactions leading to the formation of dinitrogen oxide and nitrogen and photolytic fission of dinitrogen oxide. Addition of sodium nitrite resulted in the rapid formation of dinitrogen oxide. The rate of hydrogen production was not changed significantly after the addition of hydrazine or nitrite. The results indicate that under normal operating conditions no dangerous hydrogen radiolysis yields should develop in the course of reprocessing and high-level liquid waste tank storage. Organic entrainment may lead to enhanced radiolytic decomposition and thus to considerable hydrogen production rates and pressure build-up in closed systems.

The Radiolytic Gas Production Rate in Boiling Water Reactors

The radiolytic gas production rate in boiling water reactors has been reevaluated. The results of recent measurements of the dissolved gases in the main steam samples agree very well with the avera...

Routine production of reactive fluorine-18 fluoride salts from an oxygen-18 water target

The use of a conventional small volume [ 18 O]water target gave rise to two problems. First, the water tendedto be expelled from the target by radiolytic gas production and second, continued use of the target gave [ 18 F]fluoride which could not be used for organic synthesis due to metal ions dissolving in the target water. Both of these problems have been overcome by using a silver target with a head space that permits the gas to leave the target without removing water.

Laboratory Investigation into the Radiolytic Gas Generation from Rock Salt. A Study Related to the Disposal of High Level Radioactive Waste

ABSTRACTSalt samples of two different mineralogical compositions were subjected to 60Co-γ-irradiation under an air-atmosphere. The resulting gaseous products were analysed from the gas phase above the salt. Additionally, the salt was subsequently heated up to 300 °C in order to liberate adsorbed, less volatile, and polar compounds. The gases CO2, CO, N2O, H2S, SO2, and Cl2 were identified whereas H2 was notably absent. The influence of various parameters, i. e. the total absorbed dose, the dose rate, and the temperature, on the radiolytic gas production was studied in some detail, increasing dose leads to increasing yields in CO2 and N2O. Carbon monoxide is radiolytically destroyed. Since CO2 and CO occur naturally in rock salt, they desorb thermally to some extent during the irradiation. The dose rate does not affect the yields, while the temperature during irradiation has a big effect on the radiolytic CO2 yields. At 250 °C and a radiation dose of 1×106 Gy a maximum CO2 yield of 70 mg gas per kg irradiated salt was observed. Upon heating the sample to 300 °C for 30 min. 47 mg per kg salt are additionally released.

Calculation of the Radiolytic Gas Production in Cemented Waste

ABSTRACTHydrogen production by radiolysis of water in cement waste forms has to be limited due to safety requirements. Specific generation rates (g-values) are derived from data obtained in laboratory scale experiments and measurements on 200 1 drums containing real waste. To interpret the data model calculations are carried out including those for PH, dose rate, nitrate concentration and diffusion. An OH depleting reaction is added to the radiolytical reaction mechanism of water to model the hydrogen production rate observed for γ-irradiated cement.

Radiolytic gas production from tritiated water adsorbed on molecular sieve 5A.

Hydrogen evolution from 60Co γ- and 3H β-irradiated molecular sieve 5A adsorbing HTO was measured. The results showed an additional evolution of H2(HT) gas due to energy transfer from molecular sieve to adsorbed water in both γ- and β-irradiations, and the yield at a given irradiation dose was higher than that from radiolysis of liquid HTO but lower than that in the silica gel-HTO system. The additional yield was formulated as a function of the amount of adsorbed water and irradiation dose. However a large difference in hydrogen yields was observed between γ- and β-irradiations contrary to silica gel-HTO system. A possible reason is considered.