JPH0481159B2 - - Google Patents
Info
- Publication number
- JPH0481159B2 JPH0481159B2 JP60047379A JP4737985A JPH0481159B2 JP H0481159 B2 JPH0481159 B2 JP H0481159B2 JP 60047379 A JP60047379 A JP 60047379A JP 4737985 A JP4737985 A JP 4737985A JP H0481159 B2 JPH0481159 B2 JP H0481159B2
- Authority
- JP
- Japan
- Prior art keywords
- nuclear fuel
- pellets
- cladding tube
- cut
- particles
- 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 - Lifetime
Links
- 239000008188 pellet Substances 0.000 claims description 26
- 238000005253 cladding Methods 0.000 claims description 25
- 239000002915 spent fuel radioactive waste Substances 0.000 claims description 13
- 239000002245 particle Substances 0.000 claims description 11
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 claims description 8
- 229910017604 nitric acid Inorganic materials 0.000 claims description 8
- 239000003758 nuclear fuel Substances 0.000 claims description 7
- 239000002002 slurry Substances 0.000 claims description 5
- 238000004090 dissolution Methods 0.000 claims description 4
- 238000000227 grinding Methods 0.000 claims description 2
- 238000000034 method Methods 0.000 description 10
- 239000000446 fuel Substances 0.000 description 7
- 239000000463 material Substances 0.000 description 4
- 238000010309 melting process Methods 0.000 description 4
- 238000012958 reprocessing Methods 0.000 description 4
- 238000000926 separation method Methods 0.000 description 4
- 238000002844 melting Methods 0.000 description 3
- 230000008018 melting Effects 0.000 description 3
- 238000002156 mixing Methods 0.000 description 3
- 238000010008 shearing Methods 0.000 description 3
- 241000196324 Embryophyta Species 0.000 description 2
- 229910052778 Plutonium Inorganic materials 0.000 description 2
- 229910052770 Uranium Inorganic materials 0.000 description 2
- 229910003460 diamond Inorganic materials 0.000 description 2
- 239000010432 diamond Substances 0.000 description 2
- 238000005516 engineering process Methods 0.000 description 2
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 description 2
- 238000012545 processing Methods 0.000 description 2
- 238000011084 recovery Methods 0.000 description 2
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 2
- 241000533293 Sesbania emerus Species 0.000 description 1
- 229910001093 Zr alloy Inorganic materials 0.000 description 1
- 230000000712 assembly Effects 0.000 description 1
- 238000000429 assembly Methods 0.000 description 1
- 238000009835 boiling Methods 0.000 description 1
- 238000011109 contamination Methods 0.000 description 1
- 238000007796 conventional method Methods 0.000 description 1
- 238000010586 diagram Methods 0.000 description 1
- 238000011049 filling Methods 0.000 description 1
- 238000012423 maintenance Methods 0.000 description 1
- 229910044991 metal oxide Inorganic materials 0.000 description 1
- 150000004706 metal oxides Chemical class 0.000 description 1
- 238000010298 pulverizing process Methods 0.000 description 1
- 238000000638 solvent extraction Methods 0.000 description 1
Classifications
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
- Y02W30/00—Technologies for solid waste management
- Y02W30/50—Reuse, recycling or recovery technologies
Landscapes
- Monitoring And Testing Of Nuclear Reactors (AREA)
Description
【発明の詳細な説明】
この発明は、原子炉で使用された使用済みの核
燃料棒の内部から使用済核燃料ペレツトを取出す
とともにこれを粉砕供給するシステムに関するも
のである。DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a system for extracting spent nuclear fuel pellets from inside spent nuclear fuel rods used in a nuclear reactor, and pulverizing and supplying the spent nuclear fuel pellets.
使用済核燃料からウラン、プルトニウムを回収
するための工業的規模の再処理工場では、すでに
フランスをはじめとする各国で実証経験があり、
我国でもかなり関連技術の蓄積があるピユーレツ
クス法が適用されている。このピユーレツクス法
は、燃料中に含まれているほとんどの金属酸化物
を加熱硝酸に回分的に溶解し、溶媒抽出によりウ
ラン、プルトニウムを分離回収する方法であり、
このため溶解の前段階として、燃料ペレツトを露
出させる脱被覆工程が不可欠である。脱被覆は、
燃料集合体の解体を最小限に抑えうるものとされ
ている段状刃を使用し、燃料棒をその軸方向と直
交する方向に剪断することによつて行われてい
る。しかしながら現在の剪断技術では、剪断され
た細片の形状が不揃いであつたり、剪断面が被覆
管によつて部分的に閉塞されたりすることが起
る。このような剪断物から燃料を確実に溶解させ
るためには、溶解工程の運転条件、たとえば温
度、接触時間などを苛酷にせざるを得ず、このよ
うな条件は溶解装置の構造材料にとつても苛酷で
ある。この溶解工程の制約が、１日当りの再処理
能力に上限を与える一因ともなつている。将来的
にかなりの発生量が見込まれる高燃焼度の燃料
や、高速炉燃料の再処理にあたつては、この点に
対する充分な配慮が必要とされる。したがつて従
来の方法に代わる効果的な脱被覆方法の開発が強
く望まれている。 Industrial-scale reprocessing plants for recovering uranium and plutonium from spent nuclear fuel have already been demonstrated in France and other countries.
In Japan, the Piurex method, for which a considerable amount of related technology has been accumulated, is applied. The Piurex method is a method in which most of the metal oxides contained in fuel are dissolved in heated nitric acid in batches, and uranium and plutonium are separated and recovered by solvent extraction.
For this reason, a decoating process to expose the fuel pellets is essential as a pre-dissolution step. The uncoating is
This is done by shearing the fuel rods in a direction perpendicular to their axial direction using stepped blades that are said to be able to minimize disassembly of the fuel assembly. However, current shearing techniques can result in irregularly shaped sheared strips or partial occlusion of the sheared surface by the cladding. In order to reliably melt the fuel from such sheared objects, the operating conditions of the melting process, such as temperature and contact time, must be made harsh, and these conditions also affect the structural materials of the melting equipment. It's harsh. This restriction on the melting process is also one of the reasons for placing an upper limit on the daily reprocessing capacity. Sufficient consideration must be given to this point when reprocessing high burnup fuel and fast reactor fuel, which are expected to be generated in considerable quantities in the future. Therefore, there is a strong desire to develop an effective decoating method to replace conventional methods.
この発明は以上のような要望に応じてなされた
もので、通常の剪断技術によつて被覆を切断して
その内部の核燃料ペレツトを取出すとともに、こ
のペレツトを連続的に粉砕、溶解することができ
るようなシステムを提供することを目的としてい
る。 This invention was made in response to the above-mentioned needs, and it is possible to cut the cladding and extract the nuclear fuel pellets inside the cladding using ordinary shearing technology, and to continuously crush and melt these pellets. The aim is to provide such a system.
以下にこの発明の一実施例について図面を参照
して説明する。図において符号１は、好ましくは
やや傾斜して設けられたベンチを示し、このベン
チ１上に、処理すべき使用済核燃料棒２が適当
数、たとえば５〜10本載置される。この発明が対
象としている核燃料棒は、ジルカロイのような材
料からなる被覆管内に、たとえば直径約１cm、長
さ約３cmの円柱状のペレツトを充填し、被覆管の
両端に上部端栓および下部端栓を取付けた構造を
有するもので、上部端栓とペレツトとの間に、ペ
レツトの移動を防止するためのプレナムスプリン
グが設けられている。このような構造の核燃料棒
は、使用にあたつて多数集合された集合体を構成
しているので、脱被覆処理に先立つて個々に分離
され、ベンチ１上に載置される。 An embodiment of the present invention will be described below with reference to the drawings. In the figure, reference numeral 1 indicates a bench, which is preferably installed at a slight inclination, on which a suitable number of spent nuclear fuel rods 2 to be treated, for example 5 to 10, are placed. The nuclear fuel rod to which this invention is directed is made by filling a cladding tube made of a material such as Zircaloy with cylindrical pellets, for example, about 1 cm in diameter and about 3 cm in length, and having upper end plugs and lower end plugs at both ends of the cladding tube. It has a structure in which a plug is attached, and a plenum spring is provided between the upper end plug and the pellet to prevent the pellet from moving. Nuclear fuel rods having such a structure constitute a large number of assemblies in use, and are therefore individually separated and placed on the bench 1 prior to the decladding process.
ベンチ１上に載置された複数の使用済核燃料棒
のうち、最前端に位置するものについて、レーザ
あるいはダイヤモンドカツタを切断手段とする切
断装置（図示せず）を使用して、上部端栓部およ
び下部端栓部の切除と、被覆管の切断とが行われ
る。上部および下部端栓部の切除は、充填された
ペレツト列の両端にそれぞれ対応する位置で被覆
管をその軸心と直交する平面に沿つて切断するこ
とで容易に行うことができるが、切断位置がずれ
ると、ペレツトの一部も除去され、または端栓や
プレナムスプリングの一部が切断されて残存する
ことになるので、このような事態を避けるため
に、切断に先立つて核燃料棒の軸方向の位置を微
調整することが望まれる。このようにして上部お
よび下部端栓部が切除されたのち、被覆管が軸方
向に複数個所で切断される。最も一般的には、被
覆管は、軸心をはさんで相互に対向する２つの位
置で、その一端から他端まで軸方向に延びる直線
に沿つて切断され、この場合には、被覆管は半円
形の横断面を有する２つの部分に分割される。 Among the plurality of spent nuclear fuel rods placed on the bench 1, the top end plug of the foremost one is cut using a cutting device (not shown) using a laser or a diamond cutter as a cutting means. Then, the lower end plug is removed and the cladding tube is cut. The upper and lower end plugs can be easily removed by cutting the cladding tube along a plane perpendicular to its axis at positions corresponding to both ends of the filled pellet row, but the cutting positions If the fuel rod is displaced, part of the pellet will also be removed, or part of the end plug or plenum spring will be cut off and remain. It is desirable to finely adjust the position of. After the upper and lower end plugs are removed in this manner, the cladding tube is cut at multiple locations in the axial direction. Most commonly, the cladding tube is cut along a straight line extending axially from one end to the other at two mutually opposite locations across the axis; in this case, the cladding tube is Divided into two parts with semicircular cross section.
端栓部の切除と被覆管の切断を終えた核燃料棒
２は、ベンチ１の下方に設けられた分離装置３に
落とされる。この分離装置３は、ペレツトは通過
させるが、軸方向に切断された被覆管は通過させ
ない程度の大きさの開口を有するメツシユと、こ
のメツシユを振動させるために必要に応じて設け
られた振動装置とからなる。軸方向に切断された
被覆管は、ベンチ１から分離装置３のメツシユ上
に落下したときに受ける程度の衝撃でも容易にペ
レツトを放出し、この分離されたペレツトはメツ
シユを通過してホツパー４内に落ちる。またメツ
シユ上に残つた被覆管は適当な回収装置（図示せ
ず）で回収されるが、この被覆管は上記のような
一定寸法の細長いものであるので、その回収なら
びに以後の取扱いは容易であ。なお分離装置３へ
の落下の衝撃だけではペレツトの分離が不充分で
ある場合には、切断された被覆管を強制的に剥離
するなどの補助手段を設けてもよい。 The nuclear fuel rod 2 whose end plug has been removed and whose cladding tube has been cut is dropped into a separation device 3 provided below the bench 1. This separation device 3 includes a mesh having an opening large enough to allow the pellet to pass through but not to allow the cladding tube cut in the axial direction to pass through, and a vibration device provided as necessary to vibrate the mesh. It consists of. The cladding tube cut in the axial direction easily releases pellets even with the impact received when it falls from the bench 1 onto the mesh of the separator 3, and the separated pellets pass through the mesh into the hopper 4. fall into. In addition, the cladding tube remaining on the mesh is recovered by a suitable recovery device (not shown), but since the cladding tube is elongated with a certain size as described above, its recovery and subsequent handling are easy. a. Note that if the impact of dropping onto the separating device 3 alone is insufficient to separate the pellets, auxiliary means such as forcibly peeling off the cut cladding tube may be provided.
ホツパー４に投入されたペレツトは、粉砕機５
に連続的に供給され、適当な粒度、たとえば10〜
100メツシユの粒度の粒子に粉砕される。粉砕機
５としては、一般に用いられている任意の形式の
ものを使用できるが、好ましい形式は、コーヒー
豆の粉砕に用いられているような回転ミルであ
り、被覆管の破片などの異物が混入している場合
の故障を避けるために、回転刃が可撓性物質で構
成されているものがさらに好ましい。核燃料棒の
切断によるペレツトの分離は間欠的に行われる
が、分離されたペレツトはいつたんホツパー４に
収容され、粉砕機５にはペレツトは連続的に供給
され、したがつて粉砕機５からの粒子の排出も連
続的に行われる。 The pellets fed into the hopper 4 are sent to the crusher 5
is continuously fed to a suitable particle size, e.g. 10~
It is ground into particles with a particle size of 100 mesh. As the grinder 5, any commonly used type can be used, but a preferred type is a rotary mill such as the one used for grinding coffee beans, which prevents contamination by foreign matter such as pieces of cladding. It is further preferable that the rotary blade be made of a flexible material in order to avoid failure when the blade is in use. Separation of pellets by cutting nuclear fuel rods is performed intermittently, but the separated pellets are stored in the hopper 4 and the pellets are continuously supplied to the crusher 5. Particle discharge is also carried out continuously.
粉砕機５から連続的に排出された粒子は、希硝
酸が収容されている混合槽６内に導入される。こ
の混合槽６内では、粉砕機５から排出された粒子
と希硝酸とが撹拌機７の作用で混合され、適当な
濃度のスラリーが形成される。そしてこのスラリ
ーは、所定の流量で連続的に取出されたのち、加
熱された、好ましくは沸とうしている濃硝酸が収
容されている溶解槽８内に供給される。供給され
たスラリーに含まれている使用済核燃料の粒子
は、熱濃硝酸に溶解され、この溶液が所定の流量
で連続的に取出されたのち、公知の再処理工程に
送られる。溶解槽８内に供給される使用済核燃料
は小径の粒子の形態であり、そしてすでに希硝酸
になじんでいるので、溶解は能率よく、しかも連
続的に行うことが可能である。 Particles continuously discharged from the crusher 5 are introduced into a mixing tank 6 containing dilute nitric acid. In this mixing tank 6, the particles discharged from the crusher 5 and dilute nitric acid are mixed by the action of a stirrer 7, and a slurry of an appropriate concentration is formed. After this slurry is continuously taken out at a predetermined flow rate, it is fed into a dissolution tank 8 containing heated, preferably boiling, concentrated nitric acid. Spent nuclear fuel particles contained in the supplied slurry are dissolved in hot concentrated nitric acid, and this solution is continuously taken out at a predetermined flow rate and then sent to a known reprocessing process. Since the spent nuclear fuel supplied to the melting tank 8 is in the form of small-diameter particles and has already been adapted to the dilute nitric acid, the melting can be carried out efficiently and continuously.
処理能力が4tU／日の原子炉プラント（PWR
燃料）を例にとると、処理すべき使用済核燃料棒
は2300本／日程度であり、したがつて24時間操業
を行うとすれば、１回に10本の脱被覆を行うとし
て、１本当りの処理時間として約６分間が与えら
れることになり、ベンチ１上における切断手段と
してレーザーあるいはダイヤモンドカツタを使用
すれば、この時間内での脱被覆は可能である。ま
たベンチ１および切断装置を２基並設して並行運
転すれば、脱被覆に与えられる時間は約12分間と
なり、余裕のある脱被覆処理が可能となる。また
保守などのための休止期間を考慮して、１基もし
くはそれ以上の予備を設けておくことが望ましい
場合もある。 Nuclear reactor plant (PWR) with a processing capacity of 4 tU/day
For example, the number of spent nuclear fuel rods to be processed is approximately 2,300 per day, so if the operation is carried out 24 hours a day, 10 rods will be declad at a time, and one Approximately 6 minutes is allowed for each processing time, and if a laser or diamond cutter is used as the cutting means on the bench 1, decoating is possible within this time. Furthermore, if two benches 1 and cutting devices are installed in parallel and operated in parallel, the time given for decoating will be about 12 minutes, making it possible to perform the decoating process with ample time. In addition, it may be desirable to have one or more spare units in consideration of downtime for maintenance or the like.
なお上記の実施例では、上部および下部端栓部
を切除したのちに、被覆管を軸方向に沿つて２個
所で切断する場合について説明したが、３個所で
切断すればペレツトの分離はさらに容易になり、
また適当な拡開装置を付加し、切断線に沿つて被
覆管を円周方向に拡開するように構成すれば、１
個所だけの切断でもペレツトの分離を行うことが
できる。 In the above example, the case was explained in which the cladding tube was cut at two points along the axial direction after cutting off the upper and lower end plugs, but if the cladding tube was cut at three points, it would be easier to separate the pellets. become,
Furthermore, if an appropriate expanding device is added and the cladding tube is configured to expand in the circumferential direction along the cutting line, it is possible to
Pellets can be separated by cutting only at certain points.
以上のようにこの発明によれば、使用済核燃料
棒の被覆管を軸方向に切断することによつて被覆
管とその内部のペレツトとの分離を可能にすると
ともに、この分離されたペレツトをいつたんホツ
パーに貯えてから連続的に粉砕して以後の溶解工
程に供給するようにしたので、溶解処理を高い能
率で連続的に行うことが可能になる。 As described above, according to the present invention, by cutting the cladding tube of a spent nuclear fuel rod in the axial direction, it is possible to separate the cladding tube from the pellet inside the cladding tube, and also to separate the separated pellets at any time. Since the material is stored in the hopper and then continuously pulverized and supplied to the subsequent melting process, it becomes possible to perform the melting process continuously with high efficiency.
図はこの発明の工程の一例を示す系統図であ
る。
１……ベンチ、２……使用済核燃料棒、３……
分離装置、５……粉砕機、６……混合槽、７……
撹拌機、８……溶解槽。
The figure is a system diagram showing an example of the process of this invention. 1... Bench, 2... Spent nuclear fuel rod, 3...
Separation device, 5...Crusher, 6...Mixing tank, 7...
Stirrer, 8...Dissolution tank.
Claims (1)
部端栓部分および下部端栓部分を切除したのち被
覆管を軸方向に沿つて切断し、この切断線に沿つ
て上記被覆管をその円周方向に開くことにより、
その内部に収容されているペレツトだけを取出す
操作を繰り返し、これによつて取出されたペレツ
トをいつたんホツパーに集め、このホツパーから
順次に取出したペレツトを粉砕機により適当な大
きさの粒子に粉砕し、この粒子を希硝酸と混合し
てスラリー状とし、このスラリーを、加熱された
濃硝酸が収容されている溶解槽に導入して核燃料
粒子を溶解させることを特徴とする使用済核燃料
棒の脱被覆およびペレツトの粉砕供給システム。1 Place the spent nuclear fuel rod on a bench, cut out the upper and lower end plugs, cut the cladding tube along the axial direction, and cut the cladding tube in its circle along this cutting line. By opening in the circumferential direction,
The operation of taking out only the pellets housed inside is repeated, and the pellets taken out by this are collected in a hopper, and the pellets taken out one after another from this hopper are crushed into particles of an appropriate size by a crusher. The spent nuclear fuel rods are characterized in that the particles are mixed with dilute nitric acid to form a slurry, and the slurry is introduced into a dissolution tank containing heated concentrated nitric acid to dissolve the nuclear fuel particles. Decoating and pellet grinding and feeding system.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP60047379A JPS61207999A (en) | 1985-03-12 | 1985-03-12 | Decanning of spent nuclear fuel rod and crushing supply system of pellet |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP60047379A JPS61207999A (en) | 1985-03-12 | 1985-03-12 | Decanning of spent nuclear fuel rod and crushing supply system of pellet |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS61207999A JPS61207999A (en) | 1986-09-16 |
| JPH0481159B2 true JPH0481159B2 (en) | 1992-12-22 |
Family
ID=12773458
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP60047379A Granted JPS61207999A (en) | 1985-03-12 | 1985-03-12 | Decanning of spent nuclear fuel rod and crushing supply system of pellet |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS61207999A (en) |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| FR3053151B1 (en) * | 2016-06-23 | 2018-08-10 | Commissariat A L'energie Atomique Et Aux Energies Alternatives | PROCESS FOR DISSOLVING NUCLEAR FUEL |
-
1985
- 1985-03-12 JP JP60047379A patent/JPS61207999A/en active Granted
Also Published As
| Publication number | Publication date |
|---|---|
| JPS61207999A (en) | 1986-09-16 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| US6689260B1 (en) | Nuclear fuel electrorefiner | |
| US8498371B2 (en) | Head-end process for the reprocessing of reactor core material | |
| JPH0481159B2 (en) | ||
| US4177241A (en) | Apparatus and process for recovering nuclear fuel from scrap material | |
| RU2284064C2 (en) | Method and device for removing spent nuclear fuel cladding | |
| EP1849164B1 (en) | Head-end process for the reprocessing of reactor core material | |
| CN215197381U (en) | Blockage-free slag remover | |
| FR2523156A1 (en) | ||
| JPH11148996A (en) | Molten salt electrolytic refining device | |
| JP5794932B2 (en) | Removal method of radioactive material in soil | |
| RU2658295C1 (en) | Method of decladding of fuel elements and device for its implementation | |
| JP2008530570A5 (en) | ||
| DE3007876C2 (en) | Process for comminuting spent nuclear reactor fuel elements | |
| US4230672A (en) | Apparatus for recovering nuclear fuel from scrap material | |
| JP3504595B2 (en) | Pretreatment method and pretreatment device for spent nuclear fuel | |
| DE3103900A1 (en) | "WORK PROCESS FOR SEPARATING COATED FUEL PARTICLES FROM GRAPHITIC MATRIX AND SYSTEM FOR PREPARING GRAPHITIC FUEL ELEMENTS OF HIGH-TEMPERATURE CORE REACTORS FOR A HEAD-END STAGE OF A REPROCESS" | |
| JP2012127717A (en) | Decladding method of spent fuel rod in nuclear reactor | |
| JPS63247695A (en) | Method of breaking spent nuclear fuel element | |
| JPS61250589A (en) | Method of breaking down spent nuclear fuel element | |
| JPS61250588A (en) | Processor for spent nuclear fuel element | |
| RU2229177C2 (en) | Method and process line for removing cladding from fuel element | |
| Masson et al. | HTGR spent fuel processing; The CEA investigation program | |
| CA1116410A (en) | Apparatus and process for recovering nuclear fuel from scrap material | |
| JPS61250592A (en) | Method of breaking down spent nuclear fuel element | |
| AT359141B (en) | METHOD AND DEVICE FOR CRUSHING LEAD BATTERIES |