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Abstract 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. [ABSTRACT FROM AUTHOR]

Magnesium alloys and graphite from spent nuclear fuel have been stored together in La Hague plant. The packaging of these wastes is under consideration. These wastes could be mixed in a grout composed of industrially available cement (Portland, calcium aluminate…). Within the alkaline pore solution of these matrixes, magnesium alloys are imperfectly protected by a layer of Brucite resulting in a slow corrosion releasing hydrogen. As the production of this gas must be considered for the storage safety, and the quality of wasteform, it is important to select a cement matrix capable of lowering the corrosion kinetics. Many types of calcium based cements have been tested and most of them have caused strong hydrogen production when magnesium alloys and graphite are conditioned together because of galvanic corrosion. Exceptions are binders based on alkali hydroxide activated ground granulated blast furnace slag (BFS) which are presented in this article. [ABSTRACT FROM AUTHOR]

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 H 2 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 H 2 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 H 2 release. The H 2 production of MPC materials can be reduced further by adding radical scavengers or H 2 getters within the matrix. However, other radiolytic gases such as O 2 are often produced, making these solutions potentially less attractive considering the concern of pressure build-up within the cemented waste package. [ABSTRACT FROM AUTHOR]

Ion exchange resins (IERs) are widely used by the nuclear industry to decontaminate radioactive effluents. Spent products are usually encapsulated in cementitious materials. However, the solidified waste form can exhibit strong expansion, possibly leading to cracking, if the appropriate binder is not used. In this work, the interactions between cationic resins in the Na + form and tricalcium silicate are investigated during the early stages of hydration in order to gain a better understanding of the expansion process. It is shown that the IERs exhibit a transient swelling of small magnitude due to the decrease in the osmotic pressure of the external solution. This expansion, which occurs just after setting, is sufficient to damage the material which is poorly consolidated for several reasons: low degree of hydration, precipitation of poorly cohesive sodium-bearing C–S–H, and very heterogeneous microstructure with zones of high porosity. [ABSTRACT FROM AUTHOR]