EP0149554B1 - Verfahren zum Festlegen von nuklearem Abfall - Google Patents

Verfahren zum Festlegen von nuklearem Abfall Download PDF

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Publication number
EP0149554B1
EP0149554B1 EP85300269A EP85300269A EP0149554B1 EP 0149554 B1 EP0149554 B1 EP 0149554B1 EP 85300269 A EP85300269 A EP 85300269A EP 85300269 A EP85300269 A EP 85300269A EP 0149554 B1 EP0149554 B1 EP 0149554B1
Authority
EP
European Patent Office
Prior art keywords
sulfate
weight
sodium sulfate
slurry
glass
Prior art date
Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
Expired
Application number
EP85300269A
Other languages
English (en)
French (fr)
Other versions
EP0149554A3 (en
EP0149554A2 (de
Inventor
Wilbur Orme Greenhalgh
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Westinghouse Electric Corp
Original Assignee
Westinghouse Electric Corp
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by Westinghouse Electric Corp filed Critical Westinghouse Electric Corp
Publication of EP0149554A2 publication Critical patent/EP0149554A2/de
Publication of EP0149554A3 publication Critical patent/EP0149554A3/en
Application granted granted Critical
Publication of EP0149554B1 publication Critical patent/EP0149554B1/de
Expired legal-status Critical Current

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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
    • 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
    • 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
    • G21F9/305—Glass or glass like matrix

Definitions

  • This invention relates to a method of immobilizing nuclear wates.
  • Sodium sulfate radwaste slurry is the prime liquid waste generated from boiling water reactor facilities which use bead resin cleanup.
  • the slurry is concentrated into an evaporator to about 25 wt. % and then is immobilized in cement.
  • One drum of slurry generates about three drums of final solidified waste.
  • the solidified waste is shipped to a burial site at a cost that approaches $1000 for the three drums. This situation is considered to be unsatisfactory in the industry due to the high cost involved.
  • U.S. Patent Specification No. 3 943 062 discloses the solidification of liquid nuclear waste which contains sodium or sodium compounds by calcifying in a fluidized bed.
  • U.S. Patent Specification No. 4028265 discloses a method for converting sodium nitrate containing liquid radioactive waste to a stable form by the addition of clay.
  • the present invention resides in a method of immobilizing nuclear waste in an aqueous slurry containing sodium sulfate comprising evaporating water from said slurry to produce a concentrate; characterized by mixing into said concentrate from about 50 to 200% by weight based on sodium sulfate weight of a destabilizing compound of a metal which forms an unstable sulfate; and about 5 to about 20% by weight based on the total weight of said sodium sulfate and said destabilizing compound of a reducing agent; heating at from 700 to 900°C until the evolution of sulfurous gases substantially ceases; mixing with glass formers in an amount of from 65 to 80%, based on total weight; heating to from 1050 to 1200°C; and cooling to room temperature.
  • sodium sulfate radwaste can be immobilized in glass if it is first treated to remove sulfate.
  • Sulfate removal requires the following two conditions to be present; (1) an unstable (to temperature) sulfate and (2) a reducing atmosphere or matrix.
  • Sodium sulfate of itself meets neither of these conditions as it is a stable sulfate and it is a light oxidizer.
  • the stability of the sulfate is highly dependent on the cation present. While sodium stabilizes the sulfate, it has been found that iron compounds cause instability. Therefore, by adding an iron compound together with a strong reducing agent to sodium sulfate both conditions required for removing sulfate can be met. Once the sulfate has been removed, the remaining radwaste can be combined with glass formers to form a stable glass product.
  • the compatible glass product generated from a drum of slurry using the process of this invention fills only about a third of a drum rather than the three drums that using cement would generate. As a result of this one-third reduction in waste volume, there is a tremendous savings in transportation and storage cost of the drums. Furthermore, glass immobilized waste has a lower radionuclide leach rate and a higher mechanical strength than does cement immobilized waste. For these reasons the containment of the radionuclides is safer as there is less chance of contamination with the environment.
  • the process of this invention is applicable to any sodium sulfate containing aqueous slurry.
  • the invention is particularly directed at sodium sulfate slurries containing radioactive waste that are the evaporator bottoms of a boiling water reactor. These slurries are typically about 25% (all percentages herein are by weight) sodium sulfate (based on slurry weight), although in actual practice the sodium sulfate content can vary from 15 to 40%.
  • the slurry may also contain various hydroxide, nitrate, and boric compounds. These compounds are not incompatible with the process of the invention and will aid in making a good quality glass.
  • Certain refractory type elements such as aluminium, zirconium, thorium, and the rare earths, however, should be limited to less than about 5% of the slurry solids because at higher percentages the melting temperature becomes excessive.
  • Halide compounds with the possible exception of fluoride, should be avoided in excess of 1 or 2% (based on slurry solids) as they tend to form a second glass phase.
  • these compounds are generally excluded from the reactor fluid anyway because of their corrosive nature and stainless steel piping.
  • Phosphate and carbonate compounds may also be present, but they are generally compatible with the vitrification process used in this invention.
  • the water in the sodium sulfate slurry is evaporated in a first step to less than 5% (based on the total slurry weight) in a stirrer drier to form solid granules or powder.
  • the removal of water is necessary as the presence of too much moisture could cause foam formation or solids bumping, which means that escaping steam blows the solids out of the reaction vessel.
  • the evaporation of the water can be accomplished by heating the slurry to 150 °C for as long as is necessary.
  • a destabilizing compound and a reducing agent are added to remove the sulfate.
  • the addition of a destabilizing compound and the reducing agent may be made prior to evaporation if desired.
  • the reason that sulfate must be removed when sodium is present is that sodium sulfate must be removed when sodium is present is that sodium sulfate melts without decomposing at temperatures near 880°C and the resulting liquid is non-miscible with a typical glass melt. Glass immobilization of radioactive waste requires the radionuclides and waste to be miscible with glass, and this can only occur after the sulfate fraction is removed.
  • this is accomplished by causing the formation of sulfates which are less stable than sodium sulfate, followed by the decomposition of the unstable sulfate to various sulfurous gases. This is accomplished by the addition of cations that introduce instability (along with a reducing agent).
  • the destabilizing compound is a salt of a metal which forms an unstable sulfate.
  • An unstable sulfate is one which decomposes upon heating instead of exhibiting a melting point phase change.
  • Unstable sulfates generally decompose in the 400 to 800 °C range.
  • Suitable destabilizing compounds include ferrous ammonium sulfate, ferrous sulfate, bismuth sulfate, cupric sulfate, aluminium sulfate, gallium sulfate, and manganese sulfate.
  • Ferric compounds such as ferric sulfate and ferric nitrate, can also be used if a reducing agent in an amount of about 15 to about 20% is added to reduce the ferric compound in place to the corresponding ferrous compound.
  • a reducing agent in an amount of about 15 to about 20% is added to reduce the ferric compound in place to the corresponding ferrous compound.
  • Particularly preferred is ferrous ammonium sulfate which has been found to work quite well.
  • the amount of destabilizing compounds should be from 50 to 200% of the weight of the sodium sulfate in the slurry. If less than 50% is used, all of the sulfate ion may not be destroyed. More than 200% serves no useful purpose and will simply add to the amount of waste that must be disposed of.
  • Ferric ammonium sulfate is preferably added on a one-to-one weight ratio with sodium sulfate, and graphite is added at about 10% of the total solids weight.
  • the reducing agent used should be at least as strong a reducing agent as hydrogen (Temp 400°C). Suitable reducing agents include high temperature hydrogen, dry ammonia, hydrazine, and some light hydrocarbon type amines such as methylamine, dimethylamine and trimethylamine.
  • the preferred reducing agent is carbon, especially in the form of graphite, as it has been found to work well, it is safe to use, and it reacts to produce carbon dioxide which is discharged and eliminated and, therefore, has no negative effects upon the glass product.
  • the amount of reducing agent should be from 5 to 20% based on the total weight of the sodium sulfate and the destabilizing compound. If less reducing agent is used, some of the sulfate may not be decomposed and if more is used, the glass vitrification temperature may be raised:
  • a suitable composition is from 20 to about 35% based on total composition weight, of a nuclear waste concentrate containing from 15 to 40% sodium sulfate and less than about 5% water, from 50 to 200%, based on sodium sulfate weight, of the destabilizing compound, and from 5 to 20%, based on sodium sulfate plus destabilizing compound weight, of the reducing agent.
  • the slurry concentrate is heated at from 700 to 900°C to decompose the sulfate to sulfurous gases mainly, sulfur oxide gases such as sulfur dioxide, and to force these gases out of the powder or granular solids. Heating should continue until the evolution of the sulfurous gases substantially ceases, which should not exceed eight hours.
  • glass formers are compounds routinely used to form glass such as boron oxide, and silica mixed with a glass stabilizer such as alumina or lime.
  • glass formers are compounds routinely used to form glass such as boron oxide, and silica mixed with a glass stabilizer such as alumina or lime.
  • a glass stabilizer such as alumina or lime.
  • a suitable range for a borosilicate glass composition is from 15 to 40% silica, from 20 to 40% boron trioxide, and from 1 to 5% lime or alumina (to act as a stabilizer by preventing the glass from fracturing after vitrification during cooling), and from 20 to 35% of the waste.
  • a borosilicate glass consisting of about 33% boron trioxide, about 31% silica, and about 2% alumina or lime, mixed with about 33% of the waste concentrate is preferred.
  • the mixture is heated to the melting temperature of the glass, which is typically from 1050 to 1200°C. Below 1050 °C a homogeneous glass melt may not be achieved, and therefore a poor quality glass or ceramic may result. Higher glass melting temperatures could be used if suitable containers can be found. This temperature is maintained until a homogeneous glass melt is obtained. Generally, about two hours are required to produce a homogeneous product; shorter melting times may result in an inhomogeneous glass melt and therefore a poor product. Longer vitrification times, up to eight hours, are acceptable and are limited only by economics and the corrosion of the container. The melt should be annealed by allowing it to cool gradually to room temperature.
  • melt can be poured into containers which are insulated so that the melt cools slowly.
  • 30" deep stainless steel can of glass a minimum annealing time of 4 hours is typical and a maximum annealing time would be 24 hours.
  • the cold glass can then be packaged in drums, or etc. and be transported to storage facilities.
  • a slurry containing 80 grams of sodium sulfate and 210 grams of water was mixed with 100 grams of ferrous ammonium sulfate and 40 grams of graphite and was treated as in Example 1 except the sulfate removal time was 2 hours instead of 4 hours.
  • To ten grams of the calcine mix was added 10 grams of silica, 10 grams of boron trioxide, and a gram of lime. This mix was vitrified at 1100°C to form a good quality glass product.

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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Chemical & Material Sciences (AREA)
  • Inorganic Chemistry (AREA)
  • Processing Of Solid Wastes (AREA)

Claims (6)

1. Ein Verfahren zur Immobilisierung von nuklearem Abfall in einer wässrigen Aufschlämmung, die Natrium-Sulfat enthält, bestehend aus Verdampfen von Wasser von der Aufschlämmung zur Erzeugung eines Konzentrats; gekennzeichnet durch Einmischen von etwa 50 bis 200 Gew.- %, basierend auf dem Natrium-Sulfat-Gewicht, einer destabilisierenden Verbindung eines Metalls, welches ein unstabiles Sulfat bildet; und von etwa 5 bis etwa 20 Gew.-%, basierend auf dem Gesamtgewicht des Natrium-Sulfats und der destabilisierenden Verbindung, eines reduzierenden Mittels, in das Konzentrat; Erhitzen bei 700 bis 900°C, bis die Entwicklung von schwefligen Gasen im wesentlichen aufhört; Vermischen mit Glasbildern in einer Menge von 65 bis 80%, basierend auf dem Gesamtgewicht; Erhitzen auf 1050 bis 1200°C; und Abkühlen auf Raumtemperatur.
2. Ein Verfahren nach Anspruch 1, dadurch gekennzeichnet, dass die destabilisierende Verbindung und die reduzierende Verbindung in die Aufschlämmung eingemischt werden, bevor vom letzteren Wasser verdampft wird.
3. Ein Verfahren nach Anspruch 1 oder 2, dadurch gekennzeichnet, dass das Wasser in der Aufschlämmung auf weniger als 5 Gew.-% des Gewichtes der trockenen Feststoffe verdampft wird.
4. Ein Verfahren nach Anspruch 1, 2 oder 3, dadurch gekennzeichnet, dass das Konzentrat 15 bis 40 Gew.-% Natrium-Sulfat ist.
5. Ein Verfahren nach Anspruch 1, 2, 3 oder 4, dadurch gekennzeichnet, dass die destabilisierende Verbindung Ferro-Ammonium-Sulfat, schwefelsaures Eisenoxydul, schwefelsaures Eisenoxyd, salpetersaures Eisenoxyd, Wismuth-Sulfat, Kupfersulfat, Aluminium-Sulfat, Gallium-Sulfat, Mangan-Sulfat oder Mischungen davon ist.
6. Ein Verfahren nach Anspruch 1, 2, 3 oder 4, dadurch gekennzeichnet, dass das reduzierende Mittel Kohlenstoff, Wasserstoff, trockenes Ammonia, Hydrazin, oder leichte Kohlen-Wasserstoff-Amine ist.
EP85300269A 1984-01-16 1985-01-15 Verfahren zum Festlegen von nuklearem Abfall Expired EP0149554B1 (de)

Applications Claiming Priority (2)

Application Number Priority Date Filing Date Title
US57121384A 1984-01-16 1984-01-16
US571213 1984-01-16

Publications (3)

Publication Number Publication Date
EP0149554A2 EP0149554A2 (de) 1985-07-24
EP0149554A3 EP0149554A3 (en) 1985-08-28
EP0149554B1 true EP0149554B1 (de) 1988-08-24

Family

ID=24282776

Family Applications (1)

Application Number Title Priority Date Filing Date
EP85300269A Expired EP0149554B1 (de) 1984-01-16 1985-01-15 Verfahren zum Festlegen von nuklearem Abfall

Country Status (6)

Country Link
EP (1) EP0149554B1 (de)
JP (1) JPS60159699A (de)
KR (1) KR850005716A (de)
DE (1) DE3564635D1 (de)
ES (1) ES8702075A1 (de)
PH (1) PH22647A (de)

Families Citing this family (7)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPH0631850B2 (ja) * 1985-02-08 1994-04-27 株式会社日立製作所 放射性廃液の処理処分方法
JP2633000B2 (ja) * 1989-01-28 1997-07-23 動力炉・核燃料開発事業団 高放射性廃棄物の処理方法
JPH0695155B2 (ja) * 1990-03-15 1994-11-24 動力炉・核燃料開発事業団 高放射性廃棄物の処理方法
JP2551879B2 (ja) * 1991-06-13 1996-11-06 動力炉・核燃料開発事業団 高放射性廃棄物の減容ガラス固化処理方法
RU2156510C2 (ru) * 1998-08-11 2000-09-20 Российский федеральный ядерный центр-Всероссийский научно-исследовательский институт экспериментальной физики - РФЯЦ - ВНИИЭФ Способ обезвреживания кальцинированных радиоактивных отходов
JP6019439B2 (ja) * 2012-06-26 2016-11-02 日本碍子株式会社 放射性セシウム汚染物の処理方法
RU2643362C1 (ru) * 2017-01-16 2018-02-01 Российская Федерация, от имени которой выступает Госкорпорация "Росатом" Способ обращения с радиоактивными растворами после дезактивации поверхностей защитного оборудования

Family Cites Families (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
BE679231A (de) * 1966-04-07 1966-10-07
US4094809A (en) * 1977-02-23 1978-06-13 The United States Of America As Represented By The United States Department Of Energy Process for solidifying high-level nuclear waste
DE2835763A1 (de) * 1978-08-16 1980-02-28 Kraftwerk Union Ag Verfahren zur reinigung von tensid- und detergentienhaltigen, kontaminierten abwaessern

Also Published As

Publication number Publication date
DE3564635D1 (en) 1988-09-29
EP0149554A3 (en) 1985-08-28
PH22647A (en) 1988-10-28
ES539553A0 (es) 1986-12-01
ES8702075A1 (es) 1986-12-01
JPS60159699A (ja) 1985-08-21
KR850005716A (ko) 1985-08-28
EP0149554A2 (de) 1985-07-24

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