JPS63140997A - fuel transport container - Google Patents
fuel transport containerInfo
- Publication number
- JPS63140997A JPS63140997A JP61286643A JP28664386A JPS63140997A JP S63140997 A JPS63140997 A JP S63140997A JP 61286643 A JP61286643 A JP 61286643A JP 28664386 A JP28664386 A JP 28664386A JP S63140997 A JPS63140997 A JP S63140997A
- Authority
- JP
- Japan
- Prior art keywords
- neutron
- fuel
- shield
- transport container
- container
- 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.)
- Pending
Links
- 239000000446 fuel Substances 0.000 title claims description 21
- 229910052987 metal hydride Inorganic materials 0.000 claims description 16
- 150000004681 metal hydrides Chemical class 0.000 claims description 16
- 239000006096 absorbing agent Substances 0.000 claims description 14
- 230000005251 gamma ray Effects 0.000 claims description 12
- 239000003758 nuclear fuel Substances 0.000 claims description 8
- 230000032258 transport Effects 0.000 claims 3
- 230000000694 effects Effects 0.000 description 16
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 10
- 239000002915 spent fuel radioactive waste Substances 0.000 description 6
- 239000011358 absorbing material Substances 0.000 description 5
- 238000010521 absorption reaction Methods 0.000 description 5
- 229910052778 Plutonium Inorganic materials 0.000 description 4
- 230000004992 fission Effects 0.000 description 4
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 description 4
- 238000007789 sealing Methods 0.000 description 4
- 230000002269 spontaneous effect Effects 0.000 description 3
- 239000000126 substance Substances 0.000 description 3
- 230000000712 assembly Effects 0.000 description 2
- 238000000429 assembly Methods 0.000 description 2
- 229910052751 metal Inorganic materials 0.000 description 2
- 239000002184 metal Substances 0.000 description 2
- WJWSFWHDKPKKES-UHFFFAOYSA-N plutonium uranium Chemical compound [U].[Pu] WJWSFWHDKPKKES-UHFFFAOYSA-N 0.000 description 2
- 230000035939 shock Effects 0.000 description 2
- 239000011734 sodium Substances 0.000 description 2
- 229910001220 stainless steel Inorganic materials 0.000 description 2
- 239000010935 stainless steel Substances 0.000 description 2
- 229910000619 316 stainless steel Inorganic materials 0.000 description 1
- 229910052580 B4C Inorganic materials 0.000 description 1
- 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 1
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 description 1
- 229910001093 Zr alloy Inorganic materials 0.000 description 1
- LXQXZNRPTYVCNG-YPZZEJLDSA-N americium-241 Chemical compound [241Am] LXQXZNRPTYVCNG-YPZZEJLDSA-N 0.000 description 1
- INAHAJYZKVIDIZ-UHFFFAOYSA-N boron carbide Chemical compound B12B3B4C32B41 INAHAJYZKVIDIZ-UHFFFAOYSA-N 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 239000002131 composite material Substances 0.000 description 1
- 238000001816 cooling Methods 0.000 description 1
- 230000007423 decrease Effects 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 238000000605 extraction Methods 0.000 description 1
- 230000005253 gamme decay Effects 0.000 description 1
- 239000007788 liquid Substances 0.000 description 1
- 239000000463 material Substances 0.000 description 1
- 238000000034 method Methods 0.000 description 1
- 239000011347 resin Substances 0.000 description 1
- 229920005989 resin Polymers 0.000 description 1
- VSZWPYCFIRKVQL-UHFFFAOYSA-N selanylidenegallium;selenium Chemical compound [Se].[Se]=[Ga].[Se]=[Ga] VSZWPYCFIRKVQL-UHFFFAOYSA-N 0.000 description 1
- 229910052708 sodium Inorganic materials 0.000 description 1
- 231100000331 toxic Toxicity 0.000 description 1
- 230000002588 toxic effect Effects 0.000 description 1
- JFALSRSLKYAFGM-OIOBTWANSA-N uranium-235 Chemical compound [235U] JFALSRSLKYAFGM-OIOBTWANSA-N 0.000 description 1
- 229910052726 zirconium Inorganic materials 0.000 description 1
- QSGNKXDSTRDWKA-UHFFFAOYSA-N zirconium dihydride Chemical compound [ZrH2] QSGNKXDSTRDWKA-UHFFFAOYSA-N 0.000 description 1
Landscapes
- Fuel Cell (AREA)
Abstract
(57)【要約】本公報は電子出願前の出願データであるた
め要約のデータは記録されません。(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.
Description
【発明の詳細な説明】 〔産業上の利用分野〕 本発明は、原子炉燃料の輸送容器に関する。[Detailed description of the invention] [Industrial application field] The present invention relates to a transport container for nuclear reactor fuel.
従来の燃料輸送容器は、菊池、へ木二 ′高速炉用燃料
の輸送′Ｎα１０ 、３０ （１９８５）原子カニ業に
開示され、第３図に示すように、本体１．蓋２゜缶詰缶
３．バスケット４およびショックアブソーバ５から構成
される。本体と蓋により密封容器を構成し、燃流集合体
９を一体収納した缶詰缶が複数本バスケットに収納され
る。バスケットは缶詰缶を所定の位置に固定し、未臨界
性を保つために用いられ、その上、下部には遮蔽層が確
保されている。ショックアブソーバは容器の両端に取付
けられ、落下衝撃の緩和に用いられる。A conventional fuel transport container is disclosed in Kikuchi, Hekiji, 'Transportation of Fuel for Fast Reactors', Nα10, 30 (1985) Atomic Crab Industry, and as shown in FIG. Lid 2゜Canning can 3. It is composed of a basket 4 and a shock absorber 5. The main body and the lid constitute a sealed container, and a plurality of cans integrally containing the fuel flow assembly 9 are stored in the basket. The basket is used to hold the cans in place and maintain subcriticality, and also provides a shielding layer at the bottom. Shock absorbers are attached to both ends of the container and are used to cushion the impact of falling.
容器本体には、バスケットを取囲むようにガンマ線遮蔽
体６とその外側に中性子遮蔽体７がある。The container body has a gamma ray shield 6 surrounding the basket and a neutron shield 7 outside of the gamma ray shield 6.
ガンマ線遮蔽体には鉛、ステンレス鋼等が用いられ、中
性子遮蔽体には水、樹脂等が用いられている。本体の表
面には、円環状フィン８が取付けられており、遮蔽体を
介して伝熱される収納物の崩壊熱を大気中に放熱する助
けをする。Lead, stainless steel, etc. are used for the gamma ray shield, and water, resin, etc. are used for the neutron shield. An annular fin 8 is attached to the surface of the main body to help dissipate the decay heat of the stored items, which is transferred through the shield, into the atmosphere.
輸送容器に収納される高速炉の燃料集合体の例では、燃
料としてウラン・プルトニウム混合酸化物燃料（ＭＯＸ
燃料）を用い、原子炉の中では液体ナトリウムに冷却さ
れながら高線出力密度で照射される。このため、集合体
内の燃料ピンは、大径でジルカロイに被覆された軽水炉
燃料に比べ細径でオーステナイト系３１６ステンレス鋼
で被覆されている。また、燃料ビン内のウラン・プルト
ニウ燃料には通常２０％前後のプルトニウムが含まれて
おり、ウラン−２３５を含む核分裂性物質は軽水炉燃料
に比較して多い。これら多数本の燃料ピンが六角形状に
組み合わさって、その周囲をステンレス鋼製のラッパ管
内に収納されたものが、燃料集合体である。In the example of a fast reactor fuel assembly stored in a transport container, uranium-plutonium mixed oxide fuel (MOX) is used as the fuel.
It is irradiated with high linear power density while being cooled by liquid sodium inside the reactor. For this reason, the fuel pins in the assembly have a smaller diameter and are coated with austenitic 316 stainless steel, compared to light water reactor fuel which has a larger diameter and is coated with Zircaloy. Furthermore, the uranium-plutonium fuel in the fuel bin usually contains around 20% plutonium, and the amount of fissile material containing uranium-235 is higher than that in light water reactor fuel. A fuel assembly is a hexagonal combination of a large number of these fuel pins, which is housed in a stainless steel wrapper tube around the hexagonal shape.
高速炉の燃料笑合体は構造自体も軽水炉に比較すると大
きく変わっているが、輸送の観点から考えると、プルト
ニウムを含有していることが特徴である。プルトニウム
は毒性が強いこと、自発核分裂および（α＋ｎ）反応で
中性子を放出すること、プル１〜ニウムは−２４１が崩
壊してできるアメリシウム−２４１がガンマ線を放出す
ること等の理由で、軽水炉新燃料に比較して密封性、遮
蔽性、臨界設計に特に留意する必要がある。Although the structure of the fuel composite of a fast reactor is significantly different from that of a light water reactor, from the perspective of transportation, it is characterized by the fact that it contains plutonium. Plutonium is highly toxic, releases neutrons through spontaneous fission and (α+n) reactions, and americium-241, which is produced by the decay of plu-1 to nium-241, emits gamma rays. It is necessary to pay special attention to sealing performance, shielding performance, and criticality design compared to the above.
また、軽水炉燃料に比較して高速炉使用済燃料の特徴は
、炉内で高燃焼度まで燃やされるので。In addition, compared to light water reactor fuel, fast reactor spent fuel is unique because it is burned to a high burnup within the reactor.
使用済燃料比発熱量および使用済燃料比放射能が大きい
こと、中性子の発生が大きいこと、燃料サイクルの経済
性の観点から短冷却期間で輸送を実ｊＩすしなければな
らないため熱的に厳しくなること。It is thermally demanding because the spent fuel has a high specific calorific value and spent fuel specific radioactivity, generates a large amount of neutrons, and must be transported in a short cooling period from the economic standpoint of the fuel cycle. thing.
およびプルトニウムの含有駄が多いので密封性能が「１
（要になることが挙げられる。And since it contains a lot of plutonium, the sealing performance is ``1.
(This is important.
しかし、従来の輸送容器は中性子遮蔽体に水を用いる例
が多いが、こわを高速炉の使用済燃料集合体に使用する
と、中性子遮蔽体の厚さが増して輸送容器が大きくなる
ため、輸送における容器の取扱い、ａ遂時間等の観点か
ら遮蔽性能を高めて容器を小型化する必要がある。However, conventional transport containers often use water as a neutron shield, but if stiff water is used in fast reactor spent fuel assemblies, the thickness of the neutron shield increases and the transport container becomes larger. It is necessary to improve the shielding performance and reduce the size of the container from the viewpoints of container handling and completion time.
本発明の目的は、密封性および遮蔽性に優れた。An object of the present invention is to provide excellent sealing and shielding properties.
伝熱性の良い、燃料輸送容器を提供することにある。The purpose of the present invention is to provide a fuel transport container with good heat conductivity.
上記目的は、中性子遮蔽体を、中性子減速材である金属
水素化物と中性子吸収材とで構成させることにより達成
される。The above object is achieved by constructing the neutron shield from a metal hydride, which is a neutron moderator, and a neutron absorber.
中性子遮蔽体に金属水素化物を用いることにより、輸送
容器の強度が増し、容器破損率が低下する。この結果、
容器の密封性および遮蔽性能が増す。The use of metal hydrides in neutron shields increases the strength of transport containers and reduces the rate of container breakage. As a result,
Increases the sealing and shielding performance of the container.
また、金属水素化物は中性子減速能の大きい物質である
ため、中性子吸収能の大きい物質と共に中性子の遮蔽効
果を高める。すなわち、燃料集合体から放出される中性
子は、主に自発核分裂によるものであり、その大部分は
約１〜２ Ｍ ｅ Ｖのエネルギを持つ高速中性子であ
る。一般に、中性子吸収材である核種は、高エネルギ側
では吸収断面積は小さく、エネルギが低くなるにつれ、
１／Ｊ−こ比例して大きくなってゆく。そこで、金属水
素化物により中性子を十分に減速し、中性子吸収率を高
めることで遮蔽性能を増す。その結果。Further, since metal hydride is a substance with a large neutron moderation ability, it enhances the neutron shielding effect together with a substance with a large neutron absorption ability. That is, the neutrons emitted from the fuel assembly are mainly due to spontaneous nuclear fission, and most of them are fast neutrons with an energy of about 1 to 2 M e V. In general, nuclides that are neutron absorbers have a small absorption cross section on the high energy side, and as the energy decreases,
It increases in proportion to 1/J. Therefore, shielding performance is increased by sufficiently slowing down neutrons and increasing the neutron absorption rate using metal hydrides. the result.
中性子遮蔽体の厚さを薄くすることが可能となり、輸送
容器を小型化することができる。It becomes possible to reduce the thickness of the neutron shield, and the size of the transport container can be reduced.
更に、金属水素化物は、水に比べて伝熱性が良く、使用
済核燃料の崩壊熱除去にも有効である。Furthermore, metal hydrides have better heat conductivity than water and are effective in removing decay heat from spent nuclear fuel.
一方、金属水素化物には、ガンマ線遮蔽機能もある。す
なわち、ガンマ線には光電効果、電子対生成およびコン
プトン効果の二つの吸収過程があり、光電効果は低エネ
ルギのガンマ線吸収、電子対生成は高エネルギ、コンプ
トン効果は中間エネルギのガンマ線吸収が著しい。これ
らの吸収断面積は吸収原子の原子番号Ｚに関係し、それ
ぞれＺの五乗、Ｚの二層、Ｚにほぼ比例する６すなわち
。On the other hand, metal hydrides also have a gamma ray shielding function. That is, gamma rays have two absorption processes: photoelectric effect, electron pair generation, and Compton effect. The photoelectric effect absorbs low-energy gamma rays, the electron pair generation absorbs high-energy gamma rays, and the Compton effect absorbs intermediate-energy gamma rays. These absorption cross sections are related to the atomic number Z of the absorbing atom, respectively Z to the 5th power, Z to the 2nd power, and Z to the 6th power, approximately proportional to Z, ie.
本発明において、金属元素としてより原子番号の大きい
元素を金属水素化物に用いれば、中性子遮蔽ばかりでな
く、ガンマ線の遮蔽にも有効となり、ガンマ線遮蔽体で
ある釦を削減し、更に容器を小型化することもできる。In the present invention, if an element with a higher atomic number than the metal element is used in the metal hydride, it will be effective not only for neutron shielding but also for gamma ray shielding, reducing the number of buttons that serve as gamma ray shielding bodies and further downsizing the container. You can also.
本発明に好適な中性子減速材には、ＺｒＨｚ。A suitable neutron moderator for the present invention is ZrHz.
ＴｉＩ（２，Ｎａ、ＬｉＨなど、好適な中性子吸収材に
は、ＢａＣＨＥ ｕｚｏ３ｇ ’１ ａ ＨＲｅ ＋
Ｉ ｒなどがある。Suitable neutron absorbers include TiI(2, Na, LiH, BaCHE uzo3g '1 a HRe +
There are I r, etc.
以下、本発明の一実施例を第１図及び第２図により説明
する。この実施例では、容器本体ｌの遮蔽体が、内側に
鉛によるガンマ線遮蔽体６、その外側に金属水素化物に
中性子吸収材を混入した中性子遮蔽体１０からなる二層
構造をしている。An embodiment of the present invention will be described below with reference to FIGS. 1 and 2. In this embodiment, the shield of the container body l has a two-layer structure consisting of a gamma ray shield 6 made of lead on the inside and a neutron shield 10 made of metal hydride mixed with a neutron absorbing material on the outside.
次に、本発明の効果について定量的に述べる。Next, the effects of the present invention will be described quantitatively.
まず、金属水素化物と中性子吸収材との中性子遮蔽効果
であるが、この効果は、中性子吸収材と減速材の体積比
によって異なるので、その影響を調べた。炭化ボロン（
Ｂ ａ Ｃ）を中性子吸収材とし、水素化ジルコニウム
（ＺｒＨ２）を中性子減速材とした時の中性子遮蔽効果
を第４図に示す。縦軸は、遮蔽体を漏洩する中性子の個
数であり、横軸は、吸収材が全体に占める体積比である
。図から明らかなように、減速材の割合が全体積の約１
／４であれば遮蔽効果を最大にすることができる。First, we investigated the neutron shielding effect of the metal hydride and the neutron absorber, since this effect differs depending on the volume ratio of the neutron absorber to the moderator. Boron carbide (
FIG. 4 shows the neutron shielding effect when B a C) is used as a neutron absorber and zirconium hydride (ZrH2) is used as a neutron moderator. The vertical axis is the number of neutrons leaking through the shield, and the horizontal axis is the volume ratio occupied by the absorbing material. As is clear from the figure, the proportion of moderator is approximately 1% of the total volume.
/4 can maximize the shielding effect.
そこで、吸収材と減速材との比を一対三とした場合につ
いて、遮蔽体厚さを変えて漏洩中性子の個数を調べた。Therefore, we investigated the number of leaked neutrons by changing the thickness of the shield when the ratio of absorber to moderator was 1:3.
その結果を、第５図に示す。本発明では、従来例である
水を用いた中性子遮蔽体と同じ遮蔽能力を、厚さを約４
０％薄くした遮蔽体で実現することができる。The results are shown in FIG. The present invention has the same shielding ability as the conventional neutron shield using water, but with a thickness of about 4 mm.
This can be achieved with a shield made 0% thinner.
次に、金属水素化物によるガンマ線遮蔽効果であるが、
金属元素としてジルコニウム（Ｚｒ）について調べた。Next is the gamma ray shielding effect of metal hydrides.
Zirconium (Zr) was investigated as a metal element.
燃料集合体から放出されるガンマ線は、主に、自発核分
裂により放出される高エネルギ（平均５ Ｍ ａ Ｖ
）のガンマ線と、核分裂生成物のガンマ崩壊により放出
される中エネルギ（平均０．７ＭｅＶ）のガンマ線であ
る。従来例のガンマ線遮蔽体である鉛（Ｊ７；を子番号
８２）に比べてジルコニウム（原子番号４０）はガンマ
線遮蔽効果は約１／４と小さいが、中性子遮蔽体である
金属水素化物を、約１／４厚さ相当の鉛の遮蔽体として
共用することができる。Gamma rays emitted from fuel assemblies are mainly high-energy (5 M a V on average) emitted by spontaneous nuclear fission.
) and medium-energy (average 0.7 MeV) gamma rays emitted by gamma decay of fission products. Compared to lead (J7; child number 82), which is a conventional gamma ray shield, zirconium (atomic number 40) has a gamma ray shielding effect of about 1/4, which is smaller than that of lead (J7; child number 82). It can also be used as a lead shield with a thickness equivalent to 1/4.
本発明の具体的な取出し燃焼度約８０，０００Ｍ Ｗ
Ｄ／Ｔの使用済燃料の輸送容器に適用すると、外径約５
０１のバスケットの外側に位置する鉛のガンマ線遮蔽体
は、／ＩＸさが約２５Ｇから１５％薄くでき、また、更
に外側の中性子遮蔽体は、約２５■の厚さを約１５１に
できる。この結果、輸送容器本体の外径は約１５％小さ
くなり、その容積は約２５％減少し、容器を３／４に小
型化できる。Specific extraction burnup of the present invention is approximately 80,000 MW
When applied to a D/T spent fuel transport container, the outer diameter is approximately 5
The lead gamma ray shield located outside the 01 basket can be made 15% thinner with an /IX of about 25G, and the even outer neutron shield can be made from about 25mm thick to about 151. As a result, the outer diameter of the transport container body is reduced by about 15%, its volume is reduced by about 25%, and the container can be downsized by 3/4.
他の実施例を第６図、第７図に示す。Other embodiments are shown in FIGS. 6 and 7.
第６図は、輸送容器の水平断面である。遮蔽体が、内側
から、鉛、金属水素化物１１．中性子吸収材１２の三層
構造をし、金属水素化物１１中で十分に減速された中性
子を最外側の吸収材で遮蔽している。FIG. 6 is a horizontal cross-section of the shipping container. From the inside, the shield is made of lead, metal hydride, 11. It has a three-layer structure of neutron absorbing material 12, and the outermost absorbing material blocks neutrons that have been sufficiently moderated in the metal hydride 11.
第７図も輸送容器の水平断面である。金属水素化物１１
中に多数の孔を開け、そこに中性子吸収材１２を挿入す
る。この際、金属水素化物１１は中性子吸収材１２の封
入筒の役割をし、中性子吸収材１２の量の調整、吸収物
質の置換などに利用でき、遮蔽効果を変えることができ
、仕様の異なる燃料に適用できる輸送容器となる。FIG. 7 is also a horizontal cross section of the transport container. Metal hydride 11
A number of holes are made inside, and the neutron absorbing material 12 is inserted therein. At this time, the metal hydride 11 plays the role of an enclosure cylinder for the neutron absorber 12, and can be used to adjust the amount of the neutron absorber 12, replace the absorbing substance, etc., and can change the shielding effect, allowing fuel for different specifications. This is a transport container that can be applied to
本発明によれば、遮蔽体の量を低減し、輸送容器の構造
を小型化し、また、遮蔽体の強度を増すことができる。According to the present invention, the amount of the shield can be reduced, the structure of the transport container can be made smaller, and the strength of the shield can be increased.
第１図および第２図は本発明の一実施例を示す高速炉用
燃料輸送容器の垂直および水平断面図、第３図は従来例
を示す高速炉用燃料輸送容器の垂直および水平断面図、
第４図および第５図はそれぞれ本発明の効果を示す漏洩
中性子数の、吸収材体積割合および中性子遮蔽体厚さと
の関係図、第６図および第７図は本発明の他の実施例の
高速炉用燃料輸送容器の水τ１１断面回。
９・・・燃料集合体、１０・・・中性子遮蔽体、１１・
・金属水素化物、１２・・・中性子吸収材。1 and 2 are vertical and horizontal sectional views of a fast reactor fuel transport container showing an embodiment of the present invention, and FIG. 3 is a vertical and horizontal sectional view of a fast reactor fuel transport container showing a conventional example.
4 and 5 are graphs showing the relationship between the number of leaked neutrons and the volume ratio of the absorber and the thickness of the neutron shield, respectively, showing the effects of the present invention. Water τ11 cross section of fuel transport container for fast reactor. 9...Fuel assembly, 10...Neutron shield, 11.
・Metal hydride, 12...neutron absorber.
Claims (1)
体からなる燃料輸送容器において、 前記中性子遮蔽体が中性子減速材である金属水素化物と
中性子吸収材とで構成されることを特徴とする燃料輸送
容器。[Scope of Claims] 1. A fuel transport container that transports nuclear fuel and is composed of a gamma ray shield and a neutron shield, wherein the neutron shield is composed of a metal hydride that is a neutron moderator and a neutron absorber. A fuel transport container characterized by:
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61286643A JPS63140997A (en) | 1986-12-03 | 1986-12-03 | fuel transport container |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61286643A JPS63140997A (en) | 1986-12-03 | 1986-12-03 | fuel transport container |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPS63140997A true JPS63140997A (en) | 1988-06-13 |
Family
ID=17707078
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP61286643A Pending JPS63140997A (en) | 1986-12-03 | 1986-12-03 | fuel transport container |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS63140997A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| EP0930620A1 (en) * | 1997-12-24 | 1999-07-21 | GNS GESELLSCHAFT FÜR NUKLEAR-SERVICE mbH | Storage container for the intermediate and/or final storage of spent fuel assemblies |
-
1986
- 1986-12-03 JP JP61286643A patent/JPS63140997A/en active Pending
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| EP0930620A1 (en) * | 1997-12-24 | 1999-07-21 | GNS GESELLSCHAFT FÜR NUKLEAR-SERVICE mbH | Storage container for the intermediate and/or final storage of spent fuel assemblies |
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