JPH0262033B2 - - Google Patents
Info
- Publication number
- JPH0262033B2 JPH0262033B2 JP13435184A JP13435184A JPH0262033B2 JP H0262033 B2 JPH0262033 B2 JP H0262033B2 JP 13435184 A JP13435184 A JP 13435184A JP 13435184 A JP13435184 A JP 13435184A JP H0262033 B2 JPH0262033 B2 JP H0262033B2
- Authority
- JP
- Japan
- Prior art keywords
- radioactive
- zirconium oxide
- sodium
- waste
- mixture
- 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
Links
- 239000011734 sodium Substances 0.000 claims description 25
- 239000000463 material Substances 0.000 claims description 24
- 150000003388 sodium compounds Chemical class 0.000 claims description 24
- RVTZCBVAJQQJTK-UHFFFAOYSA-N oxygen(2-);zirconium(4+) Chemical compound [O-2].[O-2].[Zr+4] RVTZCBVAJQQJTK-UHFFFAOYSA-N 0.000 claims description 22
- 239000002901 radioactive waste Substances 0.000 claims description 22
- 229910001928 zirconium oxide Inorganic materials 0.000 claims description 22
- 239000000203 mixture Substances 0.000 claims description 15
- 229910001093 Zr alloy Inorganic materials 0.000 claims description 13
- 230000002285 radioactive effect Effects 0.000 claims description 13
- 239000002699 waste material Substances 0.000 claims description 13
- 238000000034 method Methods 0.000 claims description 11
- 238000007496 glass forming Methods 0.000 claims description 10
- 238000002156 mixing Methods 0.000 claims description 7
- 238000005253 cladding Methods 0.000 claims description 6
- 239000002915 spent fuel radioactive waste Substances 0.000 claims description 6
- 238000001816 cooling Methods 0.000 claims description 4
- 238000007711 solidification Methods 0.000 claims description 3
- 230000008023 solidification Effects 0.000 claims description 3
- 238000004017 vitrification Methods 0.000 claims description 2
- 229910052708 sodium Inorganic materials 0.000 description 18
- DGAQECJNVWCQMB-PUAWFVPOSA-M Ilexoside XXIX Chemical compound C[C@@H]1CC[C@@]2(CC[C@@]3(C(=CC[C@H]4[C@]3(CC[C@@H]5[C@@]4(CC[C@@H](C5(C)C)OS(=O)(=O)[O-])C)C)[C@@H]2[C@]1(C)O)C)C(=O)O[C@H]6[C@@H]([C@H]([C@@H]([C@H](O6)CO)O)O)O.[Na+] DGAQECJNVWCQMB-PUAWFVPOSA-M 0.000 description 16
- HEMHJVSKTPXQMS-UHFFFAOYSA-M Sodium hydroxide Chemical compound [OH-].[Na+] HEMHJVSKTPXQMS-UHFFFAOYSA-M 0.000 description 12
- VYPSYNLAJGMNEJ-UHFFFAOYSA-N Silicium dioxide Chemical compound O=[Si]=O VYPSYNLAJGMNEJ-UHFFFAOYSA-N 0.000 description 8
- 238000002844 melting Methods 0.000 description 7
- 230000008018 melting Effects 0.000 description 7
- 230000001590 oxidative effect Effects 0.000 description 5
- 239000000446 fuel Substances 0.000 description 4
- 238000010438 heat treatment Methods 0.000 description 4
- 238000002386 leaching Methods 0.000 description 4
- 239000007788 liquid Substances 0.000 description 4
- 239000000377 silicon dioxide Substances 0.000 description 4
- 235000012239 silicon dioxide Nutrition 0.000 description 4
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 4
- 229910004298 SiO 2 Inorganic materials 0.000 description 3
- 230000000694 effects Effects 0.000 description 3
- 239000011521 glass Substances 0.000 description 3
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 description 2
- 239000007864 aqueous solution Substances 0.000 description 2
- 230000000712 assembly Effects 0.000 description 2
- 238000000429 assembly Methods 0.000 description 2
- 239000003795 chemical substances by application Substances 0.000 description 2
- 238000005260 corrosion Methods 0.000 description 2
- 230000007797 corrosion Effects 0.000 description 2
- JKWMSGQKBLHBQQ-UHFFFAOYSA-N diboron trioxide Chemical compound O=BOB=O JKWMSGQKBLHBQQ-UHFFFAOYSA-N 0.000 description 2
- 238000011038 discontinuous diafiltration by volume reduction Methods 0.000 description 2
- 230000005496 eutectics Effects 0.000 description 2
- 230000002779 inactivation Effects 0.000 description 2
- 229910052751 metal Inorganic materials 0.000 description 2
- 239000002184 metal Substances 0.000 description 2
- 239000000941 radioactive substance Substances 0.000 description 2
- 238000012958 reprocessing Methods 0.000 description 2
- 239000000126 substance Substances 0.000 description 2
- CDBYLPFSWZWCQE-UHFFFAOYSA-L Sodium Carbonate Chemical compound [Na+].[Na+].[O-]C([O-])=O CDBYLPFSWZWCQE-UHFFFAOYSA-L 0.000 description 1
- 229910006501 ZrSiO Inorganic materials 0.000 description 1
- 239000010426 asphalt Substances 0.000 description 1
- 229910052810 boron oxide Inorganic materials 0.000 description 1
- 239000004568 cement Substances 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 230000000052 comparative effect Effects 0.000 description 1
- 239000012141 concentrate Substances 0.000 description 1
- 239000000470 constituent Substances 0.000 description 1
- 230000004992 fission Effects 0.000 description 1
- 239000000155 melt Substances 0.000 description 1
- 239000010814 metallic waste Substances 0.000 description 1
- 239000000843 powder Substances 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 230000000717 retained effect Effects 0.000 description 1
- 239000006104 solid solution Substances 0.000 description 1
- 229910052726 zirconium Inorganic materials 0.000 description 1
Classifications
-
- C—CHEMISTRY; METALLURGY
- C03—GLASS; MINERAL OR SLAG WOOL
- C03B—MANUFACTURE, SHAPING, OR SUPPLEMENTARY PROCESSES
- C03B5/00—Melting in furnaces; Furnaces so far as specially adapted for glass manufacture
- C03B5/005—Melting in furnaces; Furnaces so far as specially adapted for glass manufacture of glass-forming waste materials
-
- C—CHEMISTRY; METALLURGY
- C03—GLASS; MINERAL OR SLAG WOOL
- C03B—MANUFACTURE, SHAPING, OR SUPPLEMENTARY PROCESSES
- C03B5/00—Melting in furnaces; Furnaces so far as specially adapted for glass manufacture
- C03B5/02—Melting in furnaces; Furnaces so far as specially adapted for glass manufacture in electric furnaces, e.g. by dielectric heating
- C03B5/023—Melting in furnaces; Furnaces so far as specially adapted for glass manufacture in electric furnaces, e.g. by dielectric heating by microwave heating
Landscapes
- Chemical & Material Sciences (AREA)
- Engineering & Computer Science (AREA)
- Materials Engineering (AREA)
- Organic Chemistry (AREA)
- Processing Of Solid Wastes (AREA)
Description
＜産業上の利用分野＞
この発明は、高速炉で使用されて不要となつた
放射性廃ナトリウムや使用済燃料再処理施設より
発生する放射性廃液のうちナトリウムを多量に含
むもの等をガラス固化処理する方法に関するもの
である。
＜従来の技術＞
従来、高速炉で使用されて不要となつた放射性
廃ナトリウム等を処理するに際しては、ナトリウ
ムを安定なナトリウム化合物に転換したのたセメ
ント、アスフアルト等と混合して固化体とする方
法が採られていた。しかしながら、この方法で
は、減容比が小さく、また得られた固化体を地中
や水中に埋めた場合に固化体内に封じ込めた放射
性ナトリウムが浸出しやすいという欠点があつ
た。
減容比を比較的大とするために、放射性廃ナト
リウム化合物に二酸化珪素やフリツト等のガラス
形成材を加えて加熱溶融したのち冷却、固化して
ガラス固化体とする技術が提案されているが、こ
の場合にも、得られたガラス固化体から放射性ナ
トリウムが浸出しやすいという欠点は必ずしも解
消することができない。
一方、軽水炉の燃料被覆管や燃料集合体の種々
の金属部材は耐食性、耐高温水性を備えたジルコ
ニウム合金であるジルカロイが用いられている
が、使用済燃料の処理に際して排出されるかよう
な放射性ジルカロイは高レベル金属廃棄物である
ため、その処理対策が問題とされている。
＜発明が解決しようとする問題点＞
そこでこの発明は、放射性廃ナトリウム化合物
をガラス固化処理してナトリウムの浸出しにくい
ガラス固化体を得ることができると同時に、従来
からその処理対策が望まれていた使用済燃料被覆
管等の放射性ジルカロイ廃棄物の処理も有効に行
なうことができる処理方法を提供することを目的
としてなされたものである。
＜問題点を解決するための手段＞
すなわちこの発明は、使用済核燃料被覆管等の
放射性ジルカロイ廃棄物を酸化処理して酸化ジル
コニウムとし、放射性廃ナトリウム化合物をガラ
ス形成材および前記酸化ジルコニウムと混合し、
この混合物を加熱溶融したのち冷却、固化するこ
とを特徴とする放射性廃ナトリウム化合物のガラ
ス固化処理方法である。
＜発明の具体的説明＞
本発明者等は、放射性廃ナトリウム化合物を二
酸化珪素（SiO2）やフリツト等のガラス形成材
と混合溶融してガラス固化体を作る際に、酸化ジ
ルコニウム（ZrO2）をガラス形成材と共に添加
混合することによつて、潮解性がなく耐浸出性に
優れた安定なガラス固化体が得られること、更に
は、上記で使用する酸化ジルコニウムとして、使
用済燃料被覆管等を構成する放射性ジルカロイ廃
棄物を酸化して得られる放射性の酸化ジルコニウ
ムを効果的に使用できることを見出し、この発明
を完成させたものである。
この発明の被処理物である放射性廃ナトリウム
化合物としては、NaOH，Na2O，Na2CO3，
NaNO3等を主成分とする固体または水溶液であ
る。かような放射性廃ナトリウム化合物の発生源
は、例えば高速炉からの放射性廃ナトリウムや、
軽水炉再処理施設からの含ナトリウム廃液等が上
げられる。前者の廃ナトリウムはそのままでは化
学的活性が高いため、NaOH等の安定なナトリ
ウム化合物とする不活性化工程が必要となるが、
後者の含ナトリウム廃液の場合には廃液中でナト
リウムが既に安定なナトリウム化合物となつてい
るため不活性化工程は必要ない。
放射性廃ナトリウム化合物が水溶液の場合に
は、蒸発処理して所定濃度まで濃縮する前処理が
必要となる。
また、被処理物である上記の如き放射性廃ナト
リウム化合物中には、ナトリウム化合物以外の放
射性物質（例えば炉構成材料の腐食生成物や燃料
物質の核分裂生成物等）が一般に同伴して含まれ
ているが、これらの同伴放射性物質もナトリウム
化合物と共にこの発明の処理方法によつてガラス
固化体中に安定に保持することができる。
この発明で用いるガラス形成材としては、二酸
化珪素やフリツト等の従来から放射性廃棄物の固
化処理技術分野で慣用されているガラス形成材を
使用することができるが、この発明によれば、ガ
ラス形成材に加えて酸化ジルコニウムを添加する
ことによつて、得られたガラス固化体からの放射
性ナトリウムの浸出を低減させることができる。
特にこの発明においてはこの酸化ジルコニウムと
して、使用済の燃料被覆管や燃料集合体の各種金
属部材として用いられた放射化されたジルコニウ
ム金属（ジルカロイ）を酸化処理して得た放射性
酸化ジルコニウムを使用するのである。使用済の
ジルカロイを酸化処理するには、酸化性雰囲気で
ジルカロイを900℃以上に加熱すればよく、これ
によつて酸化ジルコニウムの粉末を得ることがで
きる。
被処理物である放射性廃ナトリウム化合物は、
ガラス形成材および使用済ジルカロイからの酸化
ジルコニウムと乾式或いは湿式混合したのち、混
合物を好ましくはマイクロ波加熱により溶融し、
冷却、固化してガラス固化体とする。
好ましい混合割合は、混合物全重量に対して酸
化ジルコニウムは５〜20重量％の範囲、放射性廃
ナトリウム化合物は30重量％以下（Na2O換算）、
残部をガラス形成材とする。酸化ジルコニウムが
５重量％より少ないと得られたガラス固化体が潮
解性をもつようになり、ナトリウムの耐浸出性が
劣化し、20重量％より多いと混合物の溶融時に共
融温度が高くなり未溶解酸化ジルコニウムが残留
する傾向がある。またナトリウム化合物を30重量
％より多量に混合すると未溶解酸化ジルコニウム
が多くなつて酸化ジルコニウム添加効果が発現さ
れにくく、得られたガラス固化体が潮解性をもつ
ようになる。
混合物の加熱温度は各成分が共融する温度であ
ればよく、各成分の混合割合によつて変わるが一
般的には1100〜1500℃の範囲である。なお、必要
に応じて混合物中に酸化硼素（B2O3）等の溶融
温度降下剤を添加することにより、比較的低い温
度で溶融させることができる。
このようにして得られた溶融物を冷却、固化す
ることによつて、主としてNa2SiO3，
Na2ZrSiO5，ZrO2の単体もしくはこれらの混合
物からなるガラス固化体を得ることができる。
＜実施例＞
以下に実施例および比較例を上げて更に説明す
る。なお％は全て重量％である。
実施例 １
放射性廃ナトリウム化合物、二酸化珪素、使用
済ジルカロイを酸化して得られた酸化ジルコニウ
ムおよび溶融温度降下剤酸化硼素を下表に示した
各種割合で混合し、混合物をマイクロ波加熱溶融
炉を用いて溶融したのち冷却、固化してガラス固
化体を得た。このガラス固化体の固化状態を調べ
た結果を下表に示す。
<Industrial Application Field> This invention is used to vitrify radioactive waste sodium that is no longer needed after being used in fast reactors, and radioactive waste liquids generated from spent fuel reprocessing facilities that contain a large amount of sodium. It is about the method. <Conventional technology> Conventionally, when processing radioactive waste sodium, etc. that has been used in fast reactors and is no longer needed, the sodium is converted into a stable sodium compound, which is then mixed with cement, asphalt, etc., and solidified. method was adopted. However, this method has the disadvantage that the volume reduction ratio is small and that the radioactive sodium contained within the solidified body is likely to leach out when the solidified body is buried underground or in water. In order to achieve a relatively large volume reduction ratio, a technology has been proposed in which a glass-forming material such as silicon dioxide or frit is added to the radioactive waste sodium compound, heated and melted, and then cooled and solidified to form a vitrified material. In this case as well, the disadvantage that radioactive sodium tends to leach out from the vitrified material obtained cannot necessarily be overcome. On the other hand, Zircaloy, a zirconium alloy with corrosion resistance and high temperature water resistance, is used for various metal parts of the fuel cladding tubes and fuel assemblies of light water reactors. Since Zircaloy is a high-level metal waste, measures to deal with it are an issue. <Problems to be Solved by the Invention> Therefore, the present invention is capable of vitrifying radioactive waste sodium compounds to obtain a vitrified material from which sodium does not easily leach out, and at the same time, it solves a problem that has long been desired. The purpose of this invention is to provide a treatment method that can effectively treat radioactive Zircaloy waste such as spent fuel cladding tubes. <Means for solving the problem> That is, the present invention oxidizes radioactive zircaloy waste such as spent nuclear fuel cladding tubes to produce zirconium oxide, and mixes a radioactive waste sodium compound with a glass forming material and the zirconium oxide. ,
This is a vitrification treatment method for radioactive waste sodium compounds, which is characterized by heating and melting this mixture, and then cooling and solidifying it. <Specific Description of the Invention> The present inventors have discovered that when mixing and melting a radioactive waste sodium compound with a glass forming material such as silicon dioxide (SiO 2 ) and frit to create a vitrified body, the present inventors used zirconium oxide (ZrO 2 ). By adding and mixing zirconium with a glass forming material, a stable vitrified material with no deliquescent property and excellent leaching resistance can be obtained. The present invention was completed based on the discovery that radioactive zirconium oxide obtained by oxidizing the radioactive zircaloy waste that constitutes the material can be effectively used. The radioactive waste sodium compounds to be treated in this invention include NaOH, Na 2 O, Na 2 CO 3 ,
It is a solid or aqueous solution whose main component is NaNO 3 etc. The sources of such radioactive waste sodium compounds are, for example, radioactive waste sodium from fast reactors,
Examples include sodium-containing waste liquid from light water reactor reprocessing facilities. The former waste sodium has high chemical activity as it is, so an inactivation process is required to convert it into a stable sodium compound such as NaOH.
In the case of the latter sodium-containing waste liquid, the inactivation step is not necessary because sodium has already become a stable sodium compound in the waste liquid. If the radioactive waste sodium compound is an aqueous solution, pretreatment is required to evaporate it and concentrate it to a predetermined concentration. In addition, radioactive waste sodium compounds such as those mentioned above, which are the materials to be treated, generally contain radioactive substances other than sodium compounds (for example, corrosion products of reactor constituent materials, nuclear fission products of fuel materials, etc.). However, these accompanying radioactive substances can also be stably retained in the vitrified body together with the sodium compound by the treatment method of the present invention. As the glass forming material used in this invention, glass forming materials conventionally used in the technical field of solidification treatment of radioactive waste, such as silicon dioxide and frit, can be used. By adding zirconium oxide in addition to the material, leaching of radioactive sodium from the obtained vitrified material can be reduced.
In particular, in this invention, as the zirconium oxide, radioactive zirconium oxide obtained by oxidizing radioactive zirconium metal (zircaloy) used as various metal members of spent fuel cladding tubes and fuel assemblies is used. It is. To oxidize used Zircaloy, it is sufficient to heat it to 900° C. or higher in an oxidizing atmosphere, thereby obtaining zirconium oxide powder. The radioactive waste sodium compound that is to be treated is
After dry or wet mixing with the glass former and the zirconium oxide from the spent Zircaloy, the mixture is preferably melted by microwave heating;
Cool and solidify to form a vitrified product. The preferred mixing ratio is that zirconium oxide is in the range of 5 to 20% by weight based on the total weight of the mixture, and the radioactive waste sodium compound is in the range of 30% by weight or less (in terms of Na 2 O).
The remainder is used as a glass forming material. If the content of zirconium oxide is less than 5% by weight, the resulting vitrified material becomes deliquescent and the sodium leaching resistance deteriorates, while if it is more than 20% by weight, the eutectic temperature increases during melting of the mixture and Dissolved zirconium oxide tends to remain. Furthermore, when a sodium compound is mixed in an amount greater than 30% by weight, undissolved zirconium oxide increases, making it difficult to exhibit the effect of adding zirconium oxide, and the resulting vitrified product becomes deliquescent. The heating temperature of the mixture may be a temperature at which each component is eutectic, and is generally in the range of 1100 to 1500°C, although it varies depending on the mixing ratio of each component. Note that by adding a melting temperature lowering agent such as boron oxide (B 2 O 3 ) to the mixture as necessary, it is possible to melt the mixture at a relatively low temperature. By cooling and solidifying the melt thus obtained, mainly Na 2 SiO 3 ,
A vitrified body consisting of Na 2 ZrSiO 5 , ZrO 2 alone or a mixture thereof can be obtained. <Examples> Examples and comparative examples will be further explained below. Note that all percentages are by weight. Example 1 A radioactive waste sodium compound, silicon dioxide, zirconium oxide obtained by oxidizing spent Zircaloy, and boron oxide as a melting temperature lowering agent were mixed in various proportions shown in the table below, and the mixture was heated in a microwave heating melting furnace. After being melted by using the following methods, it was cooled and solidified to obtain a vitrified body. The results of examining the solidification state of this vitrified material are shown in the table below.
【表】
す。
この表から、酸化ジルコニウムの添加量は20％
以下、ナトリウム化合物は30％以下（Na2O換
算）で潮解性のない、すなわちナトリウムの浸出
しにくい安定なガラス固化体（ＡおよびＢ）が得
られることがわかる。
実施例 ２
下記組成の混合物を実施例１と同様にして溶
融、冷却、固化してガラス固化体とした。
廃NaOH 77.4ｇ
（Na2O換算 60ｇ）
SiO2 100ｇ
ZrO2 20ｇ
B2O3 35.6ｇ
得られたガラス固化体は潮解性もなく、透明度
も比較的高いガラス質化したもので通常のソーダ
ガラスと同様な外観を呈していた。
一方、比較のために下記組成の混合物(1)および
(2)を同様に溶融、冷却、固化して固化体を調製し
た。
(1) (2)
廃NaOH 75.5ｇ 100ｇ
(Na2O換算) (58.5ｇ) (77.5ｇ)
SiO2 75.5ｇ 200ｇ
H2O 80ml 200ml
得られた固化体は何れも潮解性を有し、ガラス
質化しなかつた。
＜効果＞
以上説明したようにこの発明の処理方法によれ
ば、放射性廃ナトリウム化合物にガラス形成材と
共に酸化ジルコニウムを添加混合して溶融、冷
却、固化することによつて、ナトリウムの浸出し
にくいガラス固化体を得ることができ、効果的な
放射性廃ナトリウム化合物の固化処理ができる。
また酸化ジルコニウムとしては使用済燃料被覆
管等の放射性ジルカロイ廃棄物を酸化処理して得
られる放射性の酸化ジルコニウムを使用するた
め、放射性ジルカロイ廃棄物の処理対策としても
有効であり、放射性廃棄物の増量防止が図れると
いう利点がある。【represent.
From this table, the amount of zirconium oxide added is 20%
It will be seen below that stable vitrified materials (A and B) with no deliquescent properties, that is, resistant to sodium leaching, can be obtained with a sodium compound content of 30% or less (in terms of Na 2 O). Example 2 A mixture having the following composition was melted, cooled, and solidified in the same manner as in Example 1 to obtain a vitrified product. Waste NaOH 77.4g (Na 2 O equivalent 60g) SiO 2 100g ZrO 2 20g B 2 O 3 35.6g The resulting vitrified material has no deliquescent properties and has relatively high transparency, making it a vitrified product similar to ordinary soda glass. It had a similar appearance. On the other hand, for comparison, mixtures (1) and
(2) was similarly melted, cooled, and solidified to prepare a solidified product. (1) (2) Waste NaOH 75.5g 100g (Na 2 O conversion) (58.5g) (77.5g) SiO 2 75.5g 200g H 2 O 80ml 200ml The solidified product obtained has deliquescent properties and is glassy. I didn't quality it. <Effects> As explained above, according to the treatment method of the present invention, by adding and mixing zirconium oxide together with a glass forming material to radioactive waste sodium compounds, melting, cooling, and solidifying the mixture, a glass that is difficult to leach out of sodium can be produced. A solidified substance can be obtained, and radioactive waste sodium compounds can be effectively solidified. In addition, as the zirconium oxide used is radioactive zirconium oxide obtained by oxidizing radioactive zircaloy waste such as spent fuel cladding, it is effective as a treatment measure for radioactive zircaloy waste and increases the amount of radioactive waste. This has the advantage of being preventable.
Claims (1)
棄物を酸化処理して酸化ジルコニウムとし、放射
性廃ナトリウム化合物をガラス形成材および前記
酸化ジルコニウムと混合し、この混合物を加熱溶
融したのち冷却、固化することを特徴とする放射
性廃ナトリウム化合物のガラス固化処理方法。 ２ 前記混合工程において、混合物全重量に対し
て酸化ジルコニウムを５〜20重量％、廃ナトリウ
ム化合物を30重量％以下（Na2O換算）、残部が
ガラス形成材となるように混合する特許請求の範
囲第１項記載の方法。[Claims] 1 Radioactive zircaloy waste such as spent nuclear fuel cladding tubes is oxidized to produce zirconium oxide, a radioactive waste sodium compound is mixed with a glass forming material and the zirconium oxide, and this mixture is heated and melted. A vitrification treatment method for radioactive waste sodium compounds, characterized by cooling and solidification. 2. In the mixing step, zirconium oxide is mixed in an amount of 5 to 20% by weight, waste sodium compound is mixed in an amount of 30% by weight or less (in terms of Na 2 O), and the remainder is a glass forming material based on the total weight of the mixture. The method described in Scope 1.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP13435184A JPS6113196A (en) | 1984-06-29 | 1984-06-29 | Glass solidifying treating method of radioactive waste sodium compound |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP13435184A JPS6113196A (en) | 1984-06-29 | 1984-06-29 | Glass solidifying treating method of radioactive waste sodium compound |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS6113196A JPS6113196A (en) | 1986-01-21 |
| JPH0262033B2 true JPH0262033B2 (en) | 1990-12-21 |
Family
ID=15126328
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP13435184A Granted JPS6113196A (en) | 1984-06-29 | 1984-06-29 | Glass solidifying treating method of radioactive waste sodium compound |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS6113196A (en) |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| BE1010854A3 (en) * | 1997-01-15 | 1999-02-02 | En Nucleaire Etablissement D U | Oxidation process at least one metal alkali. |
-
1984
- 1984-06-29 JP JP13435184A patent/JPS6113196A/en active Granted
Also Published As
| Publication number | Publication date |
|---|---|
| JPS6113196A (en) | 1986-01-21 |
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