Classifications

• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
  • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
  • G21F9/04—Treating liquids
  • G21F9/06—Processing
  • G21F9/12—Processing by absorption; by adsorption; by ion-exchange
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
  • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
  • G21F9/28—Treating solids
  • G21F9/30—Processing
  • G21F9/301—Processing by fixation in stable solid media
  • G21F9/302—Processing by fixation in stable solid media in an inorganic matrix
  • G21F9/304—Cement or cement-like matrix
• • B—PERFORMING OPERATIONS; TRANSPORTING
  • B09—DISPOSAL OF SOLID WASTE; RECLAMATION OF CONTAMINATED SOIL
  • B09B—DISPOSAL OF SOLID WASTE NOT OTHERWISE PROVIDED FOR
  • B09B3/00—Destroying solid waste or transforming solid waste into something useful or harmless
  • B09B3/20—Agglomeration, binding or encapsulation of solid waste
• • C—CHEMISTRY; METALLURGY
  • C04—CEMENTS; CONCRETE; ARTIFICIAL STONE; CERAMICS; REFRACTORIES
  • C04B—LIME, MAGNESIA; SLAG; CEMENTS; COMPOSITIONS THEREOF, e.g. MORTARS, CONCRETE OR LIKE BUILDING MATERIALS; ARTIFICIAL STONE; CERAMICS; REFRACTORIES; TREATMENT OF NATURAL STONE
  • C04B18/00—Use of agglomerated or waste materials or refuse as fillers for mortars, concrete or artificial stone; Treatment of agglomerated or waste materials or refuse, specially adapted to enhance their filling properties in mortars, concrete or artificial stone
  • C04B18/04—Waste materials; Refuse
  • C04B18/0463—Hazardous waste
• • C—CHEMISTRY; METALLURGY
  • C04—CEMENTS; CONCRETE; ARTIFICIAL STONE; CERAMICS; REFRACTORIES
  • C04B—LIME, MAGNESIA; SLAG; CEMENTS; COMPOSITIONS THEREOF, e.g. MORTARS, CONCRETE OR LIKE BUILDING MATERIALS; ARTIFICIAL STONE; CERAMICS; REFRACTORIES; TREATMENT OF NATURAL STONE
  • C04B18/00—Use of agglomerated or waste materials or refuse as fillers for mortars, concrete or artificial stone; Treatment of agglomerated or waste materials or refuse, specially adapted to enhance their filling properties in mortars, concrete or artificial stone
  • C04B18/04—Waste materials; Refuse
  • C04B18/14—Waste materials; Refuse from metallurgical processes
  • C04B18/141—Slags
  • C04B18/142—Steelmaking slags, converter slags
• • C—CHEMISTRY; METALLURGY
  • C04—CEMENTS; CONCRETE; ARTIFICIAL STONE; CERAMICS; REFRACTORIES
  • C04B—LIME, MAGNESIA; SLAG; CEMENTS; COMPOSITIONS THEREOF, e.g. MORTARS, CONCRETE OR LIKE BUILDING MATERIALS; ARTIFICIAL STONE; CERAMICS; REFRACTORIES; TREATMENT OF NATURAL STONE
  • C04B7/00—Hydraulic cements
  • C04B7/14—Cements containing slag
  • C04B7/147—Metallurgical slag
  • C04B7/153—Mixtures thereof with other inorganic cementitious materials or other activators
  • C04B7/1535—Mixtures thereof with other inorganic cementitious materials or other activators with alkali metal containing activators, e.g. sodium hydroxide or waterglass
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
  • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
  • G21F9/04—Treating liquids
  • G21F9/06—Processing
  • G21F9/16—Processing by fixation in stable solid media
  • G21F9/162—Processing by fixation in stable solid media in an inorganic matrix, e.g. clays, zeolites
  • G21F9/165—Cement or cement-like matrix

Definitions

• the present invention relates to solidification of radioactive waste.
• the present invention relates to a method in which the liquid-containing waste is solidified by mixing it with a binder and water, in order to form a mixture to be hardened, according to the preamble of Claim 1.
• the present invention also relates to a solidification product according to the preamble of Claim 16, and to a method of binding the radioactive components of an aqueous evaporator concentrate, according to Claim 20.
• Intermediate-level liquid wastes generated during the operation of a nuclear power plant can be divided into three categories: evaporator concentrates from the industrial wastewater system, ion-exchange resins used, and industrial wastewaters which comprise solids.
• the last category includes various sediments and sludges.
• the resin waste comprises ion-exchange mass which has been used to purify the waters of various systems (primary circuit, fuel pools, etc.) and which is used to maintain the chemical balance of the systems.
• the amount of such waste generated in a nuclear power plant is 5-30 m , for example approximately 8-15 m per year.
• the activity of the intermediate-level waste is generally between 1 MBq/kg and 10 GBq/kg, effective radiation shielding arrangements are required when handling the waste.
• Solidifying in concrete is a known method for handling ion-exchange resin.
• a binder in particular a hydraulic binder, such as cement, blast furnace slag or mixtures thereof, and an additive, the waste is solidified to form a hard solidification product (final disposal product).
• An example of known methods is the solution described in US Patent Specification No.
• radioactive waste from a nuclear power plant is fed into a lime-containing mixing container, to which additives are added while stirring. After that, cement and additional water are also fed into the container and the stirring is continued until the composition is uniformly mixed. The mixture thus obtained is allowed to harden.
• the solidification mixture 170-260 parts by weight of cement, 5-20 parts by weight of lime, 20-60 parts by weight of water and 0.25-10 parts by weight of hardener, are used per 100 parts by weight of waste.
• 2-20 parts by weight of an additive comprising at least two components, selected from a group consisting of sodium hydroxide, lithium carbonate and sodium silicate, are fed into the mixture.
• a significant part of the hardener is polycarboxylate.
• the amount of the achieved solidified ion-exchange resin is approximately 40-60 % of the volume of the waste package.
• the purpose of the present invention is to provide a method for solidifying radioactive waste.
• the intermediate-level ion-exchange resin used is solidified, which resin is mixed into a mixture of hydraulic binder and water, after which the resulting mixture is hardened to form a solidification product that comprises waste.
• the present invention is based on the idea that at least part of the cement generally used as the hydraulic binder is replaced with blast-furnace slag. To such a mixture is added the radioactive evaporator concentrate solution from a nuclear power plant. This in turn can replace both part of the water required for solidification and part of the alkaline activator of the blast furnace slag. At the same time, the radioactive part of the evaporator concentrate solution will be bonded to the solidification product.
• the product produced by means of the present method i.e. the "solidification product”
• the product produced by means of the present method comprises radioactive waste, which is solidified into the concrete mass, in which case the binder of the solidification product comprises a mixture of cement, blast furnace slag and alkali metal carbonate, and the waste percentage of the solidification product is more than 50 % by weight of the product. In this case, at least approximately 10 % by weight of this waste is unbound to the ion-exchange resin.
• another embodiment provides a method of binding radioactive components of the aqueous evaporator concentrate from a nuclear power plant, in which method aqueous evaporator concentrate is fed into a hardening mixture which comprises radioactive ion- exchange waste and hydraulic binder.
• the recipe used to generate a solidification product is exceptionally effective; over 50 % of the solidification product is waste, whereas in other solutions typically the waste percentage is approximately 10-20 % of the solidification product. If only ion-exchange resin is solidified into the waste package, the solidification product comprises more than 40 % of ion-exchange resin waste.
• the amount of waste in the package can be increased to more than 50 %, most suitably more than 55 %, to as much as 58-64 %.
• Another advantage of the present solution is the fact that the handling of evaporator concentrate is simplified and the amounts of separately handled wastes are reduced. As a result, the need to store liquid wastes, among others, is reduced.
• Figure 1 shows a simplified process flowchart of a method according to one embodiment
• Figure 2 is a three-dimensional drawing of the structure of five concrete bodies after submersion in water, in order to improve the assessment of water storage resistance
• Figure 3 is a graphical presentation of the elution test results for the test specimens Co-60, Cs- 134, and Cs-137 as a function of time, up to 365 days.
• Solidification product refers in the present context to a hardened body which comprises both radioactive waste and hydraulic binder, which body encloses that waste and which is suitable for the final disposal of this radioactive waste.
• the mixture is fed from the upper end into an open shell, such as a metal shell or reinforced concrete vessel, in which shell the mixture is allowed to harden in order to form a solidification product, after which a concrete lid is cast over the open end of the metal shell, in order to generate a product fit for storage.
• an open shell such as a metal shell or reinforced concrete vessel
• shell the mixture is allowed to harden in order to form a solidification product, after which a concrete lid is cast over the open end of the metal shell, in order to generate a product fit for storage.
• the product is also referred to as the“waste package”.
• intermediate-level waste refers to the waste which is generated during the operation of a nuclear power plant and the radioactivity of which is between 1 MBq/kg and 10 GBq/kg.
• This group includes the above-mentioned: evaporator concentrates coming from the industrial wastewater system, ion-exchange resins used for the separation of radionuclides, and industrial wastewaters which comprise solids, including sediments and sludges.
• ion-exchange resins are, among others, ion-exchange resins which are intended for and used for the control of primary circuit water chemistry, purification of
• decontamination solutions and, for example, purification of pool waters. These resins end up in the liquid waste storage and are solidified into the concrete mix.
• Evaporator concentrate solution is an alkaline solution which comprises boron compounds, salts and very small amounts of solids.
• the evaporator concentrate solution is the evaporation residue obtained when liquid, which is free from radioactivity, is removed, by using evaporation, from the clarified mother liquor of industrial wastewaters, sediments and sludges from a nuclear power plant.
• a “hydraulic binder” is a substance, in particular a mineral substance, which, when water is added, is capable of hardening and binding together, for example, solid particles or bodies with which it is mixed.
• a substance is referred to, the silicates and aluminates comprised in which form hydration products when the substance is hydrated, and which hydration products, over time, harden to form a uniform cementitious adhesive.
• Portland cement means an inorganic hydraulic binder based on limestone or calcium silicate. Portland cement and rapid cement are typical examples.
• Blast furnace slag is a binder which is generated in a blast furnace when calcium oxide obtained from limestone binds to itself non-metallic compounds of the gangue, such as silicates, phosphates and sulphide compounds. Blast furnace slag is generally granulated, and grinding it to a fineness of approximately 100-800 m /kg yields a binder which, in the presence of an alkaline activator, is able to act as a hydraulic binder.
• Alkaline activator means an alkaline substance, such as an alkali metal or alkaline earth metal hydroxide, carbonate, bicarbonate or a mixture thereof, or another inorganic or organic alkaline substance.
• radioactive waste is mixed into a mixture of cement, blast furnace slag and a blast furnace slag activator, in which case at least part of the water used for the solidification comprises a boron-containing evaporator concentrate solution.
• the amount of the boron-containing evaporator concentrate solution in the mixture to be hardened is at least 10 % by weight, in particular approximately 15-50 % by weight of the ion-exchange resin waste.
• the percentage of radioactive waste is more than 50 % by volume, in particular at least 55 % by volume of the solidification product, and at least 10 % by weight of the radioactive waste is waste which is unbound to the ion-exchange resin.
• the compressive strength of the hardened solidification product is greater than 5 MPa, in particular greater than approximately 10 MPa, most suitably approximately 12-18 MPa.
• evaporator concentrate solution when casting the solidification product, evaporator concentrate solution is added into the mixture of cement and blast furnace slag and a possible alkali, which evaporator concentrate solution is an alkaline solution and comprises boron compounds and salts of alkali metals and alkaline earth metals.
• an evaporator concentrate solution having a pH value of at least 10 is used.
• an evaporator concentrate solution having a pH value of at least 11 and most suitably at least 11.5 is used.
• the evaporator concentrate solution used has a pH value of at maximum
• cement and blast furnace slag in dry form is mixed, for example with horizontal pan-mixer, and the resulting mixture is dosed as a whole.
• the other residues, i.e. the alkaline activator and the water/evaporator concentrate, are dosed as separate units along a separate line directly into the waste vessel, where they are mixed.
• the substances of the solidification body are added, i.e. dosed, in stages, for example in 2-10 stages. This, for its part, makes it possible to accommodate a large amount of waste into the final disposal vessel.
• the phasing ensures the
• the evaporator concentrate is the dry residue obtained after the radioactivity-free liquid is removed by evaporation from the mother liquor of the sediments and sludges of the nuclear power plant industrial wastewaters.
• the evaporator concentrate solution used has a pH value of approximately 10-14, in particular approximately 11-13.5. Its boron percentage (as H 3 BO 3 ) is approximately
• the evaporator concentrate solution has a concentrate percentage of 5-40 %, in particular 10-30 %, most suitably 15-25 %, calculated from the weight of the solution.
• the evaporator concentrate solution used is free of Ni-63 nuclides.
• the evaporator concentrate works as an activator for the blast furnace slag.
• the blast furnace slag activator also comprises an alkaline substance, such as an alkali metal or alkaline earth metal carbonate.
• an alkaline substance such as an alkali metal or alkaline earth metal carbonate.
• at least part of the alkali metal or alkaline earth metal carbonate, which is used as the blast furnace slag activator, is added as a separate chemical, i.e. as a solid powder.
• a solidification product which comprises radioactive waste solidified into a concrete mass, the binder of which mass comprises a mixture of cement, blast furnace slag and alkali metal carbonate, in which case the waste percentage of the
• solidification product is more than 50 % by weight of the product, and in which case at least 10 % by weight of the waste is unbound to the ion-exchange resin.
• in the solidification mixture is used, per 100 parts by weight of waste, 10- 200 parts by weight of cement, 50-300 parts by weight of blast furnace slag, and 10-100 parts by weight of water, and 0.1-10 parts by weight of alkaline activator for the blast furnace slag.
• Table 2 shows the composition of a typical solidification mixture:
• the solidification product is sealed inside a shell, which is open at one end, such as a metal or reinforced concrete shell, in which case the open end is sealed with a concrete lid to form the waste package.
• a vessel that is completely open at one end is not used, but instead a vessel with an end which has or which is a plate that is equipped with holes for all products to be dosed into the waste vessel. By using such a perforated plate it is possible to prevent, or at least significantly reduce, splattering of the wet solidification product during mixing.
• the surface dose rate of the waste package formed of the solidification product is at maximum 5 mSv/h, typically approximately 0.05 mSv/h.
• the present technology is used to bind the radioactive components of the aqueous evaporator concentrate, in which case aqueous evaporator concentrate is fed into a hardening mixture which comprises radioactive ion-exchange waste and a hydraulic binder.
• the volume of evaporator concentrate is approximately 10-50 % of the total liquid volume of the mixture to be hardened.
• the purpose of the solidification process at the solidification plant is to convert the radioactive liquid wastes, i.e. evaporator concentrates, the ion-exchange resins used, and the sludges and sediments, to a final disposal form, by solidifying them into a concrete matrix, by using a process which works according to the batch principle.
• the evaporator concentrate is generated from the industrial wastewater treatment system, when the solution which is deposited on the sediments and sludges are transferred to evaporation.
• non-active condensate is evaporated from the solution, which condensate is discharged into the sea, after filtration and measurement checks.
• the remaining evaporator concentrate is an alkaline solution which comprises boron compounds, salts and very small amounts of solids.
• the evaporator concentrate is stored intermediately in a liquid waste storage facility 1. Once one container is full, small amounts of solids are precipitated on the bottom of the container and the solution formed on the top is recovered. From the solution it is possible to separate, for example caesium, by using a selective ion-exchange mass.
• the solution is pumped to a liquid waste storage, container 1. After a possible separation or purification, the radioactivity of the solution is so low that it can be pumped from the power plant into the sea, according to approved procedures, in a controlled way.
• the column which comprises ion-exchange mass, and is used for the separation of caesium, is finally disposed of in a concrete final disposal vessel.
• the power plant In addition to evaporator concentrate, the power plant generates ion-exchange resin residue.
• Ion-exchange resins are used, among others, for maintaining the water chemistry of the primary circuit and, for example, in the filters of purification plants for borated water and decontamination solutions. When the ion-exchange resins are saturated, they are not regenerated but pumped into a liquid waste storage container and solidified directly into the concrete.
• ion-exchange resin or evaporator concentrate is transferred from the liquid waste storage container 1 into the mixing container 2 in the solidification plant.
• the mixing containers for the ion-exchange resins and the evaporator concentrate are separate containers to which the waste batch to be solidified is first transferred.
• a waste batch means an amount of approximately 4 m 3 , which is transferred from the liquid waste storage facility to the solidification plant.
• the waste is mixed until it is as homogeneous as possible by circulating it in the circulation between the mixing and dosing containers.
• a sample is taken from each batch of waste to be solidified, through a sampling cycle, for radiochemical analyses and quality control of the concrete matrix, i.e. for“the preliminary test”. Solidification cannot be started on a new waste batch until the nuclear power plant laboratory has determined the nuclide-specific activity, i.e. the "waste fingerprint", from the sample, and calculated the theoretical waste dose rate, in order to evaluate whether the transferred waste meets the design criteria of the waste vessel, among others, regarding the surface dose rate.
• the surface dose rate of a finished waste package is not allowed to exceed 5 mSv/h.
• the purpose of the preliminary test is to ensure that the selected solidification recipe is functional and that the concrete-chemical properties required of the solidification product are met, that is, to ensure that the waste to be handled functions as desired in the solidification process and meets the requirements specified for it.
• the solidification can be started. If the gamma analysis or the preliminary test has determined that the dose rate limit for the waste package is exceeded or the waste will not bond, if needed, the waste can also be returned to the liquid waste storage facility and a new waste batch can be brought to the solidification plant.
• the solidification process begins by transferring the waste vessel 12, using a crane, to the starting point of the line. After that, the waste vessel is transferred along rails by means of a battery-powered transfer carriage to the solidification point. From the mixing container 2, a required amount of waste is transferred to a dosing chamber, from which the waste is dosed into the waste container 12. At the solidification point, the waste, soda (sodium carbonate or a mixture of sodium carbonate and sodium bicarbonate) which is used as an additive and as a solid, powdery alkaline activator, and a mixture of cement 4 and blast furnace slag 5, and additional water, are dosed into the waste vessel. The dosed substances are mixed using a waste vessel mixer to form a homogeneous solidification product.
• soda sodium carbonate or a mixture of sodium carbonate and sodium bicarbonate
• the ion-exchange resin waste is first mixed with an alkaline activator which will be added as a powdery chemical, after which, to the generated mixture is added, in doses, hydraulic binder and evaporator concentrate solution.
• the hydraulic binder is added in 1-20 doses, in particular 2-10 doses.
• the evaporator concentrate solution is added in 1-20 doses, in particular in 2-10 doses.
• the evaporator concentrate dosed is half of the volume of the waste vessel. This can be done, for example, if no other waste is solidified into the waste vessel.
• the evaporator concentrate which is dosed into the vessel is
• the waste vessel is transferred by means of a transfer carriage to bonding.
• the bonding reaction i.e. hydration reaction
• the bonding is monitored by measuring the temperature by using equipment installed along the solidification line. If the temperature on the surface of the waste vessel does not increase at all or increases too much, the bonding has not taken place in a desired way.
• the waste vessel can be monitored with two cameras. After the waste has been bonding for at least 24 hours, the hardened and cooled waste vessel is transferred by means of a transfer carriage to the lid-casting point 13.
• the lid can be cast when no separated water is visible on the surface of the solidification product.
• the operator makes the correct amount of lid concrete in a horizontal pan-mixer for the casting of the lid.
• the lid concrete 7, 8 flows, under gravity, through the dosing container 10 onto the waste vessel and spreads to form the lid.
• the operation of the horizontal pan-mixer for mixing the lid concrete, and the lid casting can be monitored by a camera.
• the finished waste dose is called the waste package 13.
• the waste package is transferred to the end of the track and the lid casting is allowed to dry.
• the waste package will be transferred to a monthly storage facility 14.
• the waste packages are handled from outside the monthly storage facility.
• the waste package is kept at least 28 days in the monthly storage facility, after which it can be transferred to an exit loading area onto a transport platform, and finally transported to a power station waste cave 15. All of the devices 2-11 for storage, dosing, and feeding of the powdery substances are placed above the solidification line.
• the chemical storage, blast furnace slag silo and the cement silo 4-8 with their subsystems are part of the normal process space and placed outside the monitored area, in their own silo space.
• the powdery substances are transferred by means of compressed air.
• the control room of the solidification plant is located in a radiation-free area so that the monitored area can be seen from the control room window.
• the solidification process is operated from the control room of the solidification plant, which minimises the exposure of staff to radiation.
• the maximum target number of solidifications in the solidification plant of a power plant is, for example, 200-250 solidifications per year.
• a medium-sized power plant produces approximately 40-50 m 3 of evaporator concentrate per year.
• the use of evaporator concentrate as additional water in the solidification of the ion-exchange resin, as described above, for example in at least 200 solidifications reduces the amount of evaporator concentrate to be stored by approximately 20-25 m 3 , which in practice halves the amount of evaporator concentrate generated each year.
• the test mass was flexible and reshapable. No water separated on the mass and the resin granules were evenly distributed in the mass. From a one day-old body, small pieces came off when the body was removed from the casting form, which is why the dry density value cannot be considered completely reliable. In the experiment, the temperature of the fresh mass was only slightly higher than the temperature of the raw materials, which predicted an advantageous bonding time.
• the solidification product should be reshapable and mixable. Good mixability of the mass ensures the homogeneity of the solidification product, also in the waste vessel.
• the flexibility of the solidification product is determined by using the flow test.
• the flow result of the test mass was considerably advantageous (185 mm); such a mass is neither excessively flexible, which, if it were, part of the active solidification product could splatter out of the waste vessel during mixing, nor too rigid, which makes it very workable.
• the difference between the wet and dry density of the mass is approximately 20 kg/m 3 . This is a very good value - based on that, it can be concluded that the evaporation of the moisture from the concrete is sufficiently fast.
• the compressive strength is also sufficient to meet the regulatory provision for the compression strength requirement, of a 28-day old solidification product.
• the water storage bodies made from the mass mentioned above were kept in the groundwater of the power station waste cave for nine days. After that, according to Table 4, their condition was "A”, i.e. the bodies were intact (see Figure 3). After the water storage bodies had been kept for 35 days in water, their condition was "B”, i.e. slight surface cracking appeared, but this had no influence on the functionality of the bodies.
• the bodies should withstand water storage and should not_crack, crumble or peel off.
• the solidification product is monitored in the water storage for at least 14 days.
• the power station waste cave is sealed at some point in the future, it will be filled with groundwater. It can be assumed that as the technical propagation barriers break down, no earlier than after 500 years, the groundwater of the cave will come into contact with the solidification product itself.
• Elution test specimens were prepared and sampling intervals were selected in accordance with both the ANSEANS-16.1-2003 standard and the general concrete codes. The actual samplings of the elution tests were performed according to the above standard. Three nuclides were studied, of which the total activity of the resin waste is mainly comprised of, i.e. Co-60, Cs- 134 and Cs-137. The activity from the evaporator concentrate is also taken into account in the elution test results. The results are shown in Figure 4.
• the test gives a good estimate of the water storage resistance of the body. After one year of water storage, the elution test specimen did not show, based on visual inspection, any cracks or crumbling, and it can be assumed that no cracks were formed, at least not significantly, because if cracks had appeared in the body, the elution surface area would have increased and simultaneously the elution rate, too.

Landscapes

• Engineering & Computer Science (AREA)
• Chemical & Material Sciences (AREA)
• Ceramic Engineering (AREA)
• High Energy & Nuclear Physics (AREA)
• Physics & Mathematics (AREA)
• General Engineering & Computer Science (AREA)
• Inorganic Chemistry (AREA)
• Environmental & Geological Engineering (AREA)
• Materials Engineering (AREA)
• Structural Engineering (AREA)
• Organic Chemistry (AREA)
• Civil Engineering (AREA)
• Processing Of Solid Wastes (AREA)

Applications Claiming Priority (2)

Publications (1)

Family

ID=69500772

Family Applications (1)

Country Status (3)

Families Citing this family (1)

Family Cites Families (3)

• 2018
  • 2018-12-16 FI FI20186089A patent/FI129112B/fi active IP Right Grant
• 2019
  • 2019-12-16 WO PCT/FI2019/050897 patent/WO2020128158A1/en active Application Filing
  • 2019-12-16 EP EP19848927.0A patent/EP3895185A1/de active Pending

Also Published As

Similar Documents

Legal Events