ITRM20120230A1 - DEVELOPMENT OF A SPECIAL GLAZING MATRIX WITH PARTICULAR PHYSICAL AND NUCLEAR CHEMICAL PROPERTIES FOR USE IN A SAFETY PROCESS THROUGH THE ENLARGEMENT OF AN IRRAGED NUCLEAR COMBUSTIBLE ELEMENT DAMAGED DUE TO EVENT - Google Patents

Description

DESCRIPTION

Development of a special vitreous matrix, having particular chemical-physical and nuclear properties, for use in a process of securing, by incorporating, an irradiated nuclear fuel element, damaged following an exceptional accident. Extension of the patent to the use of the special vitreous matrix for the conditioning of particular low / medium activity radioactive waste and highly toxic and noxious waste.

1) Field of reference of the technique and state of the art

The present invention relates to the management of irradiated nuclear fuel, damaged as a result of exceptional accidents, such as, for example, those that have caused the loss of coolant in a reactor or in a storage pool of the irradiated fuel. In the current state of knowledge, when a fuel element (â € œfuel assemblyâ €) presents micro cracks, due for example to corrosive phenomena, it is removed from its seat, placed in a stainless steel container (also called â € œcontainment bottleâ €, or â € œcanisterâ €), sealed with gasket and relocated to temporary storage in the pool or dry.

Such a solution, in the event of severely damaged fuel, with loss of containment represented by the metal sheath of the â € œassemblyâ €, is insufficient, for safety reasons, for long-term storage, as a any loss of containment by the aforementioned canister would expose personnel and the environment to unacceptable radiological risks, as a consequence of extensive resulting contamination.

To try to remedy this problem, and increase the safety level of the canister, some solutions have been proposed in the past, including:

- in USP 4650518 (1987) and in the more recent USP 5832392 (1998) uranium alloys are proposed as containment barrier, which involve the use of dedicated and particularly expensive technologies;

- in USP 4950426 (1990) a granular material composed of sand, bentonite and an organic liquid capable of absorbing radioactive substances is used as filler for the container; it is therefore an incoherent matrix whose barrier function does not appear to be fully effective in case of damage to the containment vessel.

The invention that will be described below constitutes a new solution proposal for the same problem, and consists in the incorporation of the entire damaged fuel element in a specifically developed low softening point glass matrix (SOGINGLASS F), which constitutes a further barrier (in addition to the external canister) against the dispersion of radioactive material; as will be detailed below, this solution is at the same time effective, safe, economical and simple to implement.

In the past, various proposals have been made to address the conditioning of radioactive waste: different matrices and different processes have been proposed, with the aim of creating an effective barrier to the release of radioactive material to the environment.

In all the patents relating to this conditioning, both for the glass compositions proposed and for the operating mode chosen - that is to say intimate mixing with radioactive waste - the process temperature is high, such as to involve the need for complicated and expensive plant solutions.

In some patents concerning the conditioning of radioactive waste, the use of glass is proposed, which is mixed with the waste at temperatures between 1150 and 1200 ° C, with the aim of immobilizing the radioactive isotopes in the glass matrix:

- in USP 4847008 (1998) a glass based on lead oxide, iron oxide and phosphate is proposed;

- in USP 6812174 B2 (2004) the use of borosilicate glasses - traditionally used for this purpose in the nuclear field - with lithium oxide additives is proposed;

- in USP 7825288 B2 (2010), with the same purposes as in the previous case, the addition of fluorine in place of lithium oxide is proposed.

The use of the low softening point glass matrix SOGINGLASS F, developed for this invention, also allows the incorporation of radioactive waste at relatively low temperatures, below 700 ° C, with all the ensuing advantages from the plant engineering point of view. , economic and safety.

2) Description of the invention

The invention consists in the development of a special vitreous matrix, hereinafter referred to as SOGINGLASS F, developed according to specific needs inherent to the particular operating conditions, linked to the process of making safe, by incorporating an irradiated nuclear fuel element into the glass matrix. , damaged following an exceptional accident.

The SOGINGLASS F glass matrix is characterized by low softening temperature, good thermal transmission, appropriate thermal expansion coefficient - in relation to the material of the fuel sheath - and â € œbarrierâ € properties after cooling.

In particular, the vitreous matrix has some peculiar characteristics, such as:

a) Softening temperature

The glass matrix SOGINGLASS F has a â € œsoftening pointâ € (temperature at which the viscosity is equal to 10 <6'6> Pa-s), which is placed in the temperature range 570 - 620 ° C, such as to ensure that the entire process object of the present invention can take place in temperature conditions that are not critical for the structural components of the fuel element.

b) Chemical composition

The glass matrix SOGINGLASS F has the following weight percentage composition of the main oxides:

PbO 35-45%

ZnO 17-22%

B20315-20%

Si0215-20%

Na20 4-8%

In order to improve specific physico-chemical properties, such as for example optimizing the adhesion of glass to the different metal alloys, chemical resistance, etc., in addition to the main oxides specified above, quantities of less than 5% of the following â € œminorâ € oxides: K20, CaO, Li20, Ce02. In order to improve some thermal (expansion coefficient) and chemical (resistance to acid / basic attack) properties, SOGINGLASS F glasses have been loaded up to 30% by weight with an inorganic fraction LAS Lithia-Alumina-Silica, which does not enter the composition of the glass.

SOGINGLASS F glass is able to guarantee at the same time an excellent resistance to dissolution in an aqueous environment and a softening temperature low enough to make it possible to incorporate the damaged fuel elements at temperatures lower than those critical for integrity. of the metal sheaths.

c) Thermal conductivity

The thermal characteristics of the SOGINGLASS F glass matrix are particularly relevant in the case of nuclear materials with significant generation of thermal energy.

The confinement system used in the implementation of the process of securing a damaged, irradiated nuclear fuel element, as better described in point 3), is based on the use of this glassy material which, initially in non-coherent solid form , Is brought beyond the softening temperature to obtain a homogeneous melt, and subsequently cooled to obtain a homogeneous solid capable of confining what it contains.

Since the thermal characteristics of each glass are strongly dependent on the temperature, and that the process described above imposes highly variable thermal conditions on the system (from room temperature to softening temperature in the first phase, and from the softening temperature to room temperature in the second phase ), A physico-mathematical model was used to calculate the thermal conductivity (phononic and radiative component) of the glass as a function of the different temperatures. The model was built starting from experimental evidence relating to the glassy material, both in massive form and in the form of incoherent grit, acquired at room temperature as well as at some other significant temperatures.

In the case of SOGINGLASS F glass, the thermal conductivity adopted was conservatively assumed to coincide with only the conducting component of the glass, equal to 0.6â € "at room temperature, m K

completely neglecting the contribution of heat transfer due to the radiation inside the glass itself.

d) Other chemical, thermal, mechanical properties

The glass matrix SOGINGLASS F meets the following minimum requirements:

Compressive strength: greater than 5 MPa (determined in accordance with Ref. (1) of Tab. 1).

Resistance to thermal cycles: in compliance with Ref. (2), after 30 thermal cycles of 24 hours from -40 ° C to 40 ° C with a relative humidity equal to 90 ± 5%, no crack is generated, and the average compressive strength is at least equal to 5 MPa, and in any case not less than 75% with respect to the initial one, determined in accordance with Ref. (1).

Consistency of the product: the speed of dissolution in water at 90 ° C, determined according to the methodology illustrated in Ref. (3), but, notwithstanding the provisions of the standard where the measurement on powdered material is required, by operating on samples monolithic with dimensions 9x4x1 mm, is less than 1 g / m day.

The references cited in the above list are collected in the following Tab.l:

Ref. (N) Type of Test Standard References Ref. (1) Compressive strength ASTM C39 / C 39M-05

Resistance to variations of

Ref. (2) UNI 11193: 2006 - § 5.2.2 temperature

Consistency tests of

Ref. (3) ASTM C1285-02 product on monolithic sample

Tab. 1: Normative references of characterization

e) Chemical resistance

In the event that the fuel element repaired by the process object of the present invention is stored in the storage pool, in the event of damage to the structure of the canister in AISI 316 steel, for example due to corrosive or accidental phenomena , the layer of SOGINGLASS F glass that surrounds the â € œassemblyâ € could come into contact with the pool water. It is therefore important that the glass is not severely attacked by water, in order to ensure that the radioactive material of the element, confined within the vitreous matrix, remains effectively isolated from the external environment. For this purpose, the SOGINGLASS F glass matrix has been specially developed to ensure a very low dissolution rate in water at 90 ° C, ranging from 1 10 <'5> Ã 4-10 <' 5> g / cm <2> day (value determined according to the method indicated in paragraph d of point 2).

f) Density

The density of SOGINGLASS F glass is between 4.1 and 4.4 g / cm3.

3) Process of securing a damaged irradiated nuclear fuel element.

The SOGINGLASS F glass matrix was designed for the safety process of a damaged irradiated nuclear fuel element described below.

The process consists of inserting the damaged irradiated nuclear fuel element into a canister and subsequent incorporation into the SOGINGLASS F glass matrix, maintaining maximum safety temperatures inside the element itself. The damaged fuel element, removed from its position in the reactor or in the storage pool, is placed - under the water head - inside a canister, which is then closed with a lid and gaskets. The canister is then transferred - after superficial decontamination - to an equipped area of the swimming pool, indicated below as conditioning and vitrification unit.

After emptying and drying with nitrogen, the canister and the interior of the â € œfuel assemblyâ € are filled with micro particles of said glass. The vitrification unit is then closed, emptied and dried. By using the radiant plates present in the unit (necessary, in particular, for fuel elements with long cooling times and to ensure optimal heating homogeneity of the entire â € œassemblyâ €), we proceed to heating from the outside of the canister, resulting in an almost uniform temperature rise.

Once the softening temperature of the glass has been exceeded, the latter reduces its viscosity to the point of allowing the total filling of the free volume of the canister and the â € œassemblyâ € with glass material, with consequent complete incorporation of the fuel.

After controlled cooling, the solidified glassy material incorporates the entire fuel element, forming, together with the stainless steel canister, a double containment barrier against radioactivity losses.

This process has been the subject of laboratory-scale experimentation and quantitative modeling with qualified calculation codes for the various stages of heating, incorporation and cooling of irradiated nuclear fuels, with thermal powers corresponding to different decay times after extraction. from the reactor.

4) Detailed description of the implementation of the invention for the safety of a damaged irradiated fuel element and related equipment, with reference to a nuclear fuel type BWR and to operations in swimming pools, under water shielding head.

By way of simplification and not of limitation, the process (and the related equipment) object of the invention can be summarized in four phases:

Phase 1: Canister the damaged fuel element

The first phase involves the removal of the damaged fuel element from the reactor (or from the storage pool) and its insertion into a stainless steel canister, (Figures I, II, III, IV of Annex 5), closed at seal with lid and gaskets.

For the sole purpose of optimizing the process, the canister may have some peculiar properties, such as the prismatic shape in the lower part (1), to obtain a uniform incorporation of the fuel element, and an upper cylindrical body (2).

However, these elements are not binding for invention, as they can be created with other geometries, methods and techniques.

In this example, the canister for fuel elements type BWR (3) consists of a body in AISI 316L stainless steel with an overall height of 5,000 mm, divided into two parts:

â € ¢ a lower square section 160 x 160 mm, for a height of approx. 4,500 mm, containing the BWR type fuel element, also of prismatic shape, with a 134x134 mm section, with an active length of 4.1 15 mm, which rests on a support on the bottom of the canister (13);

â € ¢ an upper circular section Φi = 260 mm, h = 500 mm, with flange and cover, with a useful volume equal to approximately 30% of the free volume of the prismatic body (after filling with the fuel element ), constituting a reserve volume (2) for the glass micro particles. This volume, filled with glass particles, is used to compensate for the reduction in volume during the softening and solidification phases.

The body of the entire canister is fully welded and conditioned, designed to withstand maximum operating temperatures of around 700 ° C.

The canister, always under the water head, is closed with the lid (4) using the gripping element (5), the centering pins (6) and the tightening bolts (7); the cover is equipped with a suitable gasket (8), resistant to operating temperatures and fixed to the cover itself by means of a fixing ring.

The canister with the damaged fuel element inserted (3) (type BWR in this example), is transferred, through the gripping element (5), to the decontamination unit, where the decontamination is carried out of the external walls.

After decontamination, the canister is transferred to the conditioning and vitrification unit, located in an equipped area of the same storage pool.

Similar containers can be provided for other types of nuclear fuels.

Phase 2: Preliminary conditioning of the canister

By preliminary conditioning of the canister we mean the operations of emptying the water, drying the inside of the canister and filling it with glass particles.

The containment vessel, always under the water head, is introduced into the aforementioned unit and then connected via â € œquick connectorâ € (8) and (9) - equipped with a guide system for insertion (10) - to the pressurization and emptying of the water present (not shown), using the 1â € pipe (11) for emptying. In the event of damage to the fuel sheath, the water evacuated by the canister will be heavily contaminated, and therefore will be sent for radioactive liquid treatment.

The subsequent internal drying of the canister will then be carried out, using a current of nitrogen through the emptying pipe (11) and the â € œquick connectorâ € (9). The off-gases are sent to the radioactive effluent treatment system via pipeline (8). The connections of the canister to the emptying / treatment system and for the nitrogen supply and suction of the off gases are made by means of quick connectors (â € œquick connectorâ €) and flexible pipes. The canister is then connected to a tank outside the pool and filled with glass particles, with a diameter of 100-300 pm, through a dedicated pipe and a â € œquick connectorâ € (12). The small average size (100-300 pm) of the particles is particularly important as they have to completely fill the free volume of the canister, passing through the grids of the fuel element.

The glass used in the present invention, by way of example, is the glass matrix of the SOGINGLASS F type, developed specifically for the following invention and whose characteristics have been reported in point 2).

As mentioned, the upper volume of the canister (2), equal to about 30% of its free volume, is used to compensate for the volume reduction of the glass particles following their softening.

During the filling operation, the canister remains connected to the off-gas treatment system by means of a â € œquick connectorâ €. Suitable vibrating systems will ensure the complete filling of the free volume with the glass particles.

At the end of the filling, detected by measuring the level of the glass, the canister is disconnected, remaining active only the ceramic filter which acts as a vent valve.

It should be borne in mind that, in the subsequent heating phase, the entire canister will be brought to approx. 600-700 ° C, and this obliges to use only gaskets and connectors resistant to these temperatures.

All the canister conditioning operations are carried out under the water head, in radiological safety conditions for the operating personnel, operating from the upper maneuvering deck of the storage pool.

Phase 3: Incorporation of the damaged fuel element into a glassy matrix.

The softening process of the vitreous material loaded around and inside the damaged fuel element constitutes the third phase of the process, and takes place in the same conditioning and vitrification unit (Figures V and VI - Annex 5), in the pool of deposit, under water head.

This equipment consists of an AISI 316 stainless steel cylinder (14), fitted with a cover (15), with temperature seals (16), centering pins (17) and tightening bolts (18), thermally insulated (19 ) and placed on a base (20) at the bottom of the storage pool.

of the vitrifier are positioned:

â € ¢ guides for the correct introduction and positioning of the canister (21);

â € ¢ a series of radiant electric panels (22), placed on the internal surface at different heights;

â € ¢ temperature sensors (23).

The water in the cavity between the equipment and the canister is then emptied by means of a bottom piping, and nitrogen is introduced (24), with off-gas ventilation (25); since it has not been in contact with active / contaminated parts, the drained water has the same radioactivity as the storage pool.

The canister is then heated externally by means of the radiant panels (19), placed on the internal wall of the unit.

The use of radiant panels (22) is necessary to compensate for the different axial distribution of the residual decay power, in order to optimize the entire process.

The temperature rise on the outside of the canister up to about 650 ° C is measured through the aforementioned thermal sensors (23), and the temperature is kept constant by modulating the power supplied by the radiant panels.

This modulation also takes into account the residual thermal power of the fuel element, which will have the effect of raising the temperature in the center of the element; in conditions of residual power of 3 kW (for a BWR element) corresponding to approx. 6 months of cooling of a fuel from a BWR reactor operating at a nominal power of 1200 MWe, with an average boom-up of 30,000 MWd / tonU, the temperature will reach maximum values around 720 ° C.

At these temperature levels there is a further reduction in the viscosity of the entire glass mass, which consequently will compact, drawing glassy material from the upper part of the canister.

In this hypothesis, when the power required from the radiant panels reaches almost constant values, with the maximum temperature of the canister wall of about 650 ° C, the glass mass inside the canister will have reached the melting condition. By maintaining these conditions for an adequate period of time, a generalized melting condition of the micro glass particles inside the entire canister will be achieved.

An indirect measure of the generalized melting can be carried out by measuring the level of the glass particles in the upper cylindrical body.

Phase 4: Cooling of the canister

The controlled cooling phase of the entire canister now begins, first by gradually reducing the irradiation of the radiant plates, then by introducing nitrogen into the gap between the canister and the walls of the same equipment, then with nitrogen plus water and finally with water, so as to ensure the glass conditions of integrity at the end of cooling. To avoid the cracks that could derive from the meeting of material with different expansion coefficient, a certain amount of inorganic filler with very low expansion coefficient is added to the glass, such as for example silica glass or LAS (glass ceramic of lithium alumina silica), in order to make the expansion coefficient of the glass similar to that of the Zircaloy of the sheaths of the fuel element. The AISI 316 steel of the external container, having an expansion coefficient higher than that of the materials contained within it, will induce a beneficial compression tension in the glass.

After cooling, the vitreous mass forms an integral and compact protective envelope around the same element. The unit is then open, and the canister removed and transferred, after integrity checks, to storage, pending reprocessing or long-term storage.

The process described allows the â € œreconstructionâ € of the primary containment, constituted in the original fuel element by the Zircaloy sheath of the bars, and to add an additional containment barrier compared to the use of the steel canister alone, preventing the release of radioactivity in the subsequent phases of management of the irradiated fuel, both in normal operating conditions and in accidental situations of loss of secondary containment, or damage to the canister.

In Figures V and VI of annex 5, the conditioning and vitrification unit is schematically represented, with reference to a BWR fuel; similar solutions can be envisaged to implement the process or for other types of nuclear fuels, without substantial modifications to the process object of the invention.

As mentioned, the vitrification unit consists of a metal cylinder, in AISI 316 stainless steel, equipped with a sealed lid (15) - with a diameter Φ = 380 mm and H = 5500 mm - thermally insulated, closed in an aluminum or steel casing.

The unit is immersed in the pool at its deepest point, and is anchored to a support plate.

The unit is connected at the bottom to an emptying pipe (24), by means of a submersible pump connected to the liquid effluent treatment plant and, at the top, to a suction and off gas treatment pipe (25).

Inside the cylindrical body are welded plates with a square section opening (21), for the positioning of the fuel element canister in phase.

Radiant energy thermal panels (22), axially distributed (by way of example in number of 8 in the diagram of Figure V) with electrical power supply, are positioned at various heights of the unit body.

Sensors (23) of the pyrometer type continuously measure the temperature of the external wall of the canister and allow the process to be controlled.

5) Internal temperature profiles of the canister for various decay powers

For the analysis of the temperature profiles inside the canister for various thermal powers of radioactive decay, it was developed on a canister model using a qualified finite element calculation code - ANSYS 12.1 - in which the thermophysical characteristics of the glass and the geometry of the system.

The geometry used for the simulation models of the heating and cooling processes includes:

â € ¢ a volume of homogeneous glass, inside the fuel element, within which the fuel rods are contained;

â € ¢ the fuel bars are completely assumed in U02 (conservative hypothesis);

â € ¢ the fuel element sheath is in Zircaloy;

â € ¢ a volume of glass on the outside of the fuel element sheath;

â € ¢ a steel volume relating to the vitrified prismatic part of the canister.

For a temperature outside the canister of 650 ° C, in the center of the element, there will be, as mentioned in point 1.4, a maximum temperature of 720 ° C, in the hypothesis of residual power of the fuel of 3 kW / assembly, corresponding to 6 months of cooling of a fuel of a BWR rector operating at the nominal power of 1200 MWe, with an average boom-up of 30,000 MWd / tonU

In condition of refrigeration in the canister in water, after incorporation into the glass matrix and subsequent controlled cooling, the maximum temperature value inside the fuel element is slightly higher than 310 ° C, assuming the temperature of the water in the pool of 30 ° C.

Similarly, in the case of cooling in air, it has been considered that the transfer of heat to the external environment is due both to the phenomenon of radiation and that of convection.

Conservatively, for the purposes of evaluating the contribution due to irradiation, the steel surface of the vitrified element was assumed to be polished (reduced emissivity). Despite the limited performance of the air for heat removal, the contribution due to irradiation makes it possible to obtain, during the long-term air cooling phase, a maximum temperature in the center of the element of just under 235 ° C.

6) Advantages of the process object of the invention

The present invention starts from the well-established practice of using a stainless steel containment vessel for securing cracked irradiated nuclear fuel, mainly following corrosive events.

This container or canister has been suitably modified to allow the incorporation in the SOGINGLASS F vitreous matrix of an irradiated fuel element seriously damaged as a result of exceptional accidents, such as, for example, those that involved direct exposure of the radioactive material (the irradiated tablets or pellets of U02, Pu02 and oxides of fission products) to the cooling water. Said vitreous matrix is formed at maximum temperatures around 700 ° C, in order to preserve the characteristics of the external steel canister and the Zircaloy sheaths of the fuel element.

With this process, extremely economical, simple and safe from an operational point of view, an effective containment barrier is `` rebuilt '' by incorporating the entire damaged fuel element (and any material exposed) in a glassy matrix which constitutes a barrier against the dispersion of radioactive material, both during normal operating conditions and following further accidents.

It thus adds to the walls of the canister - susceptible eg. of cracks caused by corrosive phenomena or loss of seal an additional waterproof glass barrier, with a consequent increase in the level of safety.

Furthermore, in addition to the above, the advantages of the process object of the present invention are:

- its reversibility, that is, the possibility of separating the fuel element from the encapsulating matrix, for example by heating to temperatures above the softening point;

- radiological safety during cutting operations, in case of reprocessing;

- limited increase of liquid and solid waste, in case of reprocessing of damaged fuel.

A further application of the present invention relates to the option of managing the irradiated fuel, indicated as â € œonce-through cycleâ €.

This option provides for the conditioning of the irradiated fuel in a special container (after its eventual thickening), long-term dry storage in cask or in an engineering structure, followed by final storage in a deep geological formation.

It is precisely in the â € œconditioningâ € of the fuel that this process can find its optimal application, introducing an additional barrier - the glass - between the radioactivity of the bars present inside the fuel element and the environment. .

Furthermore, since the glassy material has a higher thermal conductivity coefficient than gases, such as air or helium (for stagnant air it is approx.0.027 against approx.0.6 W / mK for glass ), there is an improvement in the disposal of the residual decay power and, in case of leakage of the external canister, the glass also keeps the fuel element protected against oxidative phenomena.

7) Use of the special glass matrix SOGINGLASS F for the conditioning of particular low and medium activity radioactive waste and highly toxic and noxious non-radioactive waste.

The SOGINGLASS F vitreous matrix developed for use in the process described here as a containment barrier for damaged fuel elements, can be optimized for use, for the purpose of incorporating low and medium activity radioactive waste, where the consolidated cementation processes, normally used for this type of waste, prove to be not or difficult to apply - as in the case of radioactive waste containing high concentrations of organic compounds - or not convenient.

In the past, various proposals have been made to address the conditioning of radioactive materials with glassy matrices: different matrices and different processes have been proposed, with the aim of creating an effective barrier to the release of radioactive material towards the environment.

In all the patents relating to this conditioning, both for the glass compositions proposed and for the operating mode chosen - that is to say intimate mixing with radioactive waste - the process temperature is high, such as to involve the need for complicated and expensive plant solutions.

In this case, the use of a glass with a low softening temperature such as SOGINGLASS F for the incorporation of the waste (after a possible treatment of concentration and dry calcination), allows to obviate the incompatibility with the matrix cementitious and at the same time to avoid the off-gas problems and the use of complex vitrification equipment, which arise in the use of glass matrices usually used in the treatment of radioactive waste.

A similar application may involve highly toxic and noxious non-radioactive waste, for which it is necessary for safety reasons, isolation in a durable and inert matrix with respect to chemical-biological environmental agents, such as the vitreous matrix SOGINGLASS F.

Claims (1)

ANNEX 3 - CLAIMS CLAIM N ° 1 Glass matrix comprising the oxides PbO, ZnO, B203, Si02, Na20, characterized by a softening temperature included in the range 570 ° C - 620 ° C, and by a viscosity (in said range) equal to IO <6> - <6> Pa s. CLAIM N ° 2 Glass matrix as per claim 1, characterized by the following chemical composition (percentages by weight of oxides): PbO 35-45% ZnO 17-22% B20315-20% Si0215-20% Na20 4-8% CLAIM N ° 3 Glass matrix as per claims 1 and 2, characterized by the addition of the following oxides in quantities lower than 5%: K20, CaO, Li20, Ce02. CLAIM N ° 4 Glass matrix as per claims 1, 2 and 3, characterized by the addition of less than 30% of the inorganic fillers LAS (Lithia-Alumina-Silica) or silica glass. CLAIM N ° 5 Glass matrix as per claims 1, 2, 3 and 4, characterized by a 5 5 2 dissolution rate in water at 90 ° C ranging from 1-10 <"> Ã · 4 10 <'> g / cm day. CLAIM N ° 6 Glass matrix as per claims 1, 2, 3, 4 and 5, characterized by a phononic component of thermal conductivity equal to 0.6 W / mK at room temperature. CLAIM N ° 7 Glass matrix as per claims 1, 2, 3, 4, 5 and 6, characterized by: Compressive strength: greater than 5 MPa Resistance to thermal cycles: after 30 thermal cycles of 24 hours from -40 ° C to 40 ° C with a relative humidity equal to 90 ± 5%, no crack is generated, and the average compressive strength is at least equal to 5 MPa, and in any case not less than 75% with respect to the initial one. Dissolution speed: in deionized water at 90 ° C, determined on monolithic samples of 9x4x1 mm size, it is less than 1 10 <"4> g / cm <2 > per day Density: between 4.1 and 4.4 g / cm <3>. CLAIM N ° 8 Use of the vitreous matrix according to claims 1, 2, 3, 4, 5, 6 and 7 for the safety of an irradiated nuclear fuel element, by incorporation that takes place inside a metal container (canister, as per fig. I - IV of Annex 5), in which the element is introduced. CLAIM N ° 9 Use of the vitreous matrix according to claim 8, in which the incorporation of the combustible element takes place inside a metallic container (canister) by filling it with the vitreous matrix in the form of solid micro particles with a diameter of 100-300 µm . CLAIM N ° I0 Conditioning and vitrification unit (as per Fig. V and VI of Annex 5), consisting of an AISI 316 stainless steel cylinder (14), fitted with a lid (15), with gaskets for high temperatures (16), centering pins (17) and tightening bolts (18), thermally insulated (19), equipped with pipes for water drainage, nitrogen blowing, off-gas suction, and ventilation, connected to a base (20) on the bottom of the storage pool and operated remotely, on whose internal surface are installed: guides for the correct introduction and positioning of the canister (21), electric radiant panels (22), placed at different heights for a homogeneous heating, temperature sensors (23), inside which the canister according to claim 8 (containing the damaged nuclear fuel element and the glass matrix microspheres) is placed, and within which the process of i embedding according to claim 9, by heating the glass matrix above its softening temperature.