JPS63225194A - Spent-fuel carrying-out method to reprocessing facility - Google Patents
Spent-fuel carrying-out method to reprocessing facilityInfo
- Publication number
- JPS63225194A JPS63225194A JP62058739A JP5873987A JPS63225194A JP S63225194 A JPS63225194 A JP S63225194A JP 62058739 A JP62058739 A JP 62058739A JP 5873987 A JP5873987 A JP 5873987A JP S63225194 A JPS63225194 A JP S63225194A
- Authority
- JP
- Japan
- Prior art keywords
- fuel
- spent fuel
- facility
- reprocessing facility
- spent
- 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.)
- Granted
Links
- 239000002915 spent fuel radioactive waste Substances 0.000 title claims description 44
- 238000012958 reprocessing Methods 0.000 title claims description 36
- 238000000034 method Methods 0.000 title claims description 15
- 239000000446 fuel Substances 0.000 claims description 42
- 238000007726 management method Methods 0.000 description 15
- 238000004090 dissolution Methods 0.000 description 12
- 238000005259 measurement Methods 0.000 description 9
- 238000002844 melting Methods 0.000 description 7
- 230000008018 melting Effects 0.000 description 7
- 230000000712 assembly Effects 0.000 description 5
- 238000000429 assembly Methods 0.000 description 5
- 238000002485 combustion reaction Methods 0.000 description 4
- 230000000694 effects Effects 0.000 description 3
- 238000010008 shearing Methods 0.000 description 3
- 238000010586 diagram Methods 0.000 description 2
- 239000002245 particle Substances 0.000 description 2
- 230000002265 prevention Effects 0.000 description 2
- 238000004886 process control Methods 0.000 description 2
- 239000007788 liquid Substances 0.000 description 1
- 238000010309 melting process 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)
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、使用済み燃料集合体の再処理施設に付帯する
使用済み燃料貯蔵施設に関し、特に該使用済み燃料貯蔵
施設から燃料を再処理施設へ搬出する方法に関するもの
である。Detailed Description of the Invention [Field of Industrial Application] The present invention relates to a spent fuel storage facility attached to a reprocessing facility for spent fuel assemblies, and in particular to a spent fuel storage facility attached to a reprocessing facility for spent fuel assemblies. This is related to the method of transporting.
［従来の技術］
原子力発電プラントの使用済み燃料ビットから取り出さ
れる使用済み燃料集合体は、再処理施設に付帯する使用
済み燃料貯蔵池もしくは貯蔵施設への搬入の段階でその
燃焼度測定が行なわれてから、前記使用済み燃料貯蔵施
設に貯蔵され、そしてこの貯蔵された使用済み燃料集合
体は、臨界防止及びプロセス管理の観点から、再処理施
設への搬入前に、各種測定原理に基づき再びその燃焼度
測定が行なわれ、再処理の前処理工程、特に溶解槽での
臨界安全性、作業性の確認を行った上で、使用済み燃料
貯蔵施設より搬出される。より具体的に述べれば、作業
者は、搬出すべき燃料の選択に際して、上述した燃焼度
測定に基づいて、搬出しようとする燃料が、再処理施設
の溶解槽における臨界防止のため核***性核種の制限濃
度を満たしうるような燃焼度の燃料かどうか、且つ前ス
テップで装荷した燃料の燃焼度を参考にして、プロセス
の濃度管理として過度に核***性核種の濃度を高くしな
いような燃焼度の燃料かどうかの判断を行っている。[Prior Art] The burnup of spent fuel assemblies taken out from spent fuel bits of nuclear power plants is measured before they are delivered to a spent fuel storage pond or storage facility attached to a reprocessing facility. After that, the spent fuel assemblies are stored in the spent fuel storage facility, and from the viewpoint of criticality prevention and process control, the spent fuel assemblies are tested again based on various measurement principles before being delivered to the reprocessing facility. Burnup measurements are performed, and criticality safety and workability in the pretreatment process for reprocessing, especially in the dissolution tank, are confirmed before the spent fuel is removed from the spent fuel storage facility. More specifically, when selecting the fuel to be transported, the operator determines whether the fuel to be transported contains fissile nuclides in order to prevent criticality in the melting tank of the reprocessing facility, based on the burnup measurement described above. Check whether the fuel has a burnup that satisfies the concentration limit, and also, by referring to the burnup of the fuel loaded in the previous step, the fuel has a burnup that does not increase the concentration of fissile nuclides excessively for concentration control in the process. I am making a judgment as to whether or not.
［発明が解決しようとする問題点コ
従って、再処理施設への搬入前に再び燃焼度の測定を行
い、この測定に基づいて種々の作業を行わねばならない
ので、そのために、長い時間を要するだけでなく、再処
理作業の円滑化が阻害されていた。[Problems to be solved by the invention] Therefore, the burnup must be measured again before delivery to the reprocessing facility, and various operations must be performed based on this measurement, which takes a long time. However, the smooth reprocessing work was hindered.
本発明は、再処理施設へ搬入する際の燃焼度測定にかか
る時間をなくし作業の円滑化を図ると共に、使用済み燃
料貯蔵施設から搬出する段階で再処理施設の臨界安全管
理、プロセス管理の作業能力を高め、再処理施設のプラ
ント稼動率を向上させる、再処理施設への使用済み燃料
搬出方法を提供することを目的とするものである。The present invention aims to streamline the work by eliminating the time required for burnup measurement when transporting spent fuel to a reprocessing facility, and also to carry out criticality safety management and process management of the reprocessing facility at the stage of transporting the spent fuel from the spent fuel storage facility. The objective is to provide a method for transporting spent fuel to a reprocessing facility that increases capacity and improves the plant operating rate of the reprocessing facility.
［問題点を解決するための手段及び作用］この目的から
、本発明による使用済み燃料搬出方法においては、再処
理施設における使用済み燃料の核種の濃度を測定し、そ
の測定値を制御ユニットの演算部に入力し、該演算部に
おいて、前記測定値と前記再処理施設における既知の核
種濃度基準値とから、前記制御ユニットのメモリー部に
記憶された使用済み燃料貯蔵施設内の使用済み燃料の核
種の軸方向重量分布に基づいて、前記使用済み燃料貯蔵
施設から前記再処理施設へ搬出される使用済み燃料を決
定し、該使用済み燃料を再処理施設へ搬出することを特
徴としている。核種の軸方向重量分布は、炉心での燃焼
管理及び燃焼により生じた核種の計量管理のために炉心
管理コードに記録されており、これを制御ユニットのメ
モリー部に入力しておくことにより、信頼性の高いこの
データを利用して、再び燃焼度を測定することなく、再
処理施設へ搬出する使用済み燃料を決定することができ
る。[Means and effects for solving the problem] For this purpose, in the spent fuel removal method according to the present invention, the concentration of nuclides in the spent fuel in the reprocessing facility is measured, and the measured value is used in the calculation of the control unit. In the calculation unit, the nuclide of the spent fuel in the spent fuel storage facility stored in the memory unit of the control unit is determined based on the measured value and the known nuclide concentration standard value in the reprocessing facility. The method is characterized in that the spent fuel to be transported from the spent fuel storage facility to the reprocessing facility is determined based on the axial weight distribution of the spent fuel, and the spent fuel is transported to the reprocessing facility. The axial weight distribution of nuclides is recorded in the core management code for combustion management in the reactor core and measurement control of the nuclides produced by combustion.By inputting this into the memory section of the control unit, reliability can be improved. Using this highly accurate data, it is possible to determine which spent fuel should be transported to a reprocessing facility without having to measure burnup again.
［実施例コ
次に、本発明の好適３実施例を添付図面に関連して詳細
に説明する。[Embodiments] Next, three preferred embodiments of the present invention will be described in detail with reference to the accompanying drawings.
第１図は本発明の使用済み燃料搬出方法を実施するシス
テムの概要を示すもので、同システムには、使用済み燃
料貯蔵池もしくは貯蔵施設Ａに関連して、通常使用され
る燃料移送装置１及びその制御ユニット２が設けられて
おり、一般的にはコンピュータである制御ユニット２は
、メモリー部２ａ、入力部２ｂ、演算部２ｃ及び制御部
２ｄを有する。FIG. 1 shows an outline of a system for carrying out the method for transporting spent fuel of the present invention. The control unit 2, which is generally a computer, has a memory section 2a, an input section 2b, a calculation section 2c, and a control section 2d.
この使用済み燃料貯蔵施設Ａは再処理施設Ｂに付設され
ている。再処理施設Ｂは周知のように、本発明に従って
燃料移送装置１により使用済み燃料貯蔵施設Ａから移送
されてきた使用済み燃料（図示せず）を所定の大きさの
小片に剪断する剪断機３と、該剪断機３によって剪断さ
れた小片を受は入れる溶解槽４とを備える。This spent fuel storage facility A is attached to the reprocessing facility B. As is well known, the reprocessing facility B has a shearer 3 for shearing the spent fuel (not shown) transferred from the spent fuel storage facility A by the fuel transfer device 1 into small pieces of a predetermined size according to the present invention. and a dissolution tank 4 that receives the small pieces sheared by the shearing machine 3.
溶解槽４においては溶解工程が実施され、その後様々な
処理工程が続くが、これ等の工程は本発明の要旨ではな
いのでその説明を省略することができる。A dissolution step is performed in the dissolution tank 4, and various treatment steps follow thereafter, but since these steps are not the gist of the present invention, their explanation can be omitted.
溶解槽４には、その内部の燃料溶解液中の核種（Ｕ、Ｐ
ｕ等）の濃度を測定するための周知のＬＬＬ４ａが設け
られており、該測定器４ａは制御ユニット２の演算部２
ｃに作動上接続されていて、測定器４ａからの核種濃度
測定値が制御ユニット２の演算部２ｃに入力されるよう
になっている。精度の向上を図るために核種濃度の測定
は適宜性なわれ、その都度新しい情報として入れ換える
ことができる。The dissolution tank 4 contains nuclides (U, P) in the fuel dissolution liquid inside it.
A well-known LLL 4a is provided for measuring the concentration of
The nuclide concentration measurement value from the measuring device 4a is input to the calculation section 2c of the control unit 2. In order to improve accuracy, nuclide concentration measurements are made as needed and can be replaced as new information each time.
入力部２ｂに入力されるデータは、炉心での燃焼管理及
び燃焼により生じた核種の計量管理を目的とする炉心管
理コードの出力として原子力発電所より或は設計メーカ
ーより磁気テープ等の媒体として得られるもので、この
データには、使用済み燃料集合体についての燃料番号（
イ）、軸方向の燃焼度分布（ロ）、核種の軸方向重量分
布（ハ）及び平均燃焼度（ニ）が含まれる。また、使用
済み燃料貯蔵施設Ａ内の燃料配置位置（ホ）も入力部２
ｂに入力されており、これ等のデータ（イ）〜（ホ）は
メモリー部２ａに記憶されていて、必要に応じて随時呼
び出すことができる。燃料番号（イ）及び燃料配置位置
（ホ）に関するデータは、搬出すべき燃料集合体が決ま
ったら、このデータに基づいて、燃料移送装置１を使用
済み燃料貯蔵施設Ａの該当する燃料集合体の位置へ移動
させ、速やかな燃料の搬出を可能とするために用いられ
る。The data input to the input unit 2b is obtained from a nuclear power plant or from a design manufacturer as a medium such as a magnetic tape as the output of a core management code for the purpose of combustion management in the reactor core and quantitative management of nuclides produced by combustion. This data includes the fuel number (
(a), axial burnup distribution (b), axial weight distribution of nuclides (c), and average burnup (d). In addition, the fuel placement position (e) in the spent fuel storage facility A is also input to the input section 2.
These data (a) to (e) are stored in the memory section 2a and can be called up at any time as needed. Once the fuel assembly to be transported is determined, the data regarding the fuel number (a) and the fuel placement position (e) are used to move the fuel transfer device 1 to the corresponding fuel assembly in the spent fuel storage facility A. It is used to move the fuel to the desired location and to enable quick fuel removal.
また、明細書の冒頭に述べたように、従来、搬出燃料の
選択に際しては、搬出しようとする燃料が、再処理施設
の溶解槽における臨界防止のため核***性核種の制限濃
度を満たしうるような燃焼度の燃料がどうか、且つ前ス
テップで装荷した燃料の燃焼度を参考にして、プロセス
の濃度管理として過度に核***性核種の濃度を高くしな
いような燃焼度の燃料かどうかの判断を行わねばならな
いが、本発明において制御の対象となるパラメータは、
臨界管理、プロセス管理の対象となる例えば溶解槽４中
の核***性核種の濃度であり、これと、炉心管理コード
により得られた核種の軸方向分布（ハ）を利用すること
により、次のステップで再処理施設へ搬送する燃料を溶
解槽４中の核***性核種の制限濃度や目標濃度、即ち基
準値に照らし合わせて決定する方法が提供される。In addition, as stated at the beginning of the specification, conventionally, when selecting fuel to be exported, it is necessary to ensure that the fuel to be exported satisfies the limit concentration of fissile nuclides in order to prevent criticality in the melting tank of the reprocessing facility. It is necessary to judge whether the fuel has a burnup that does not excessively increase the concentration of fissile nuclides for process concentration control, by referring to the burnup of the fuel loaded in the previous step. However, the parameters to be controlled in the present invention are:
This is the concentration of fissile nuclides in the melting tank 4, which is subject to criticality control and process control, and by using this and the axial distribution of nuclides (c) obtained from the core management code, the next step can be determined. A method is provided in which the fuel to be transported to the reprocessing facility is determined by comparing it with a limit concentration or target concentration of fissile nuclides in the dissolution tank 4, that is, a reference value.
さて、炉心管理コードは炉心設計の妥当性を確認し、炉
心を安全に管理するためのものであり、上述したデータ
はこの管理コードに基づいているので、その信頼性は十
分高い、従って、これ等のデータを利用すると共に、再
処理施設Ｂ内の溶解千′１１４中の核***性核種の濃度
に基づいて、再処理施設Ｂへ搬出される燃料を決定すれ
ば、溶解槽４中の核***性核種の濃度は、貯蔵施設Ａよ
り搬出され、剪断８１３に装荷され、溶解槽４へ移送さ
れる燃料、即ち剪断小片に含まれる核***性核種の重量
に依存するので、臨界防止やプロセス管理を高い信頼性
で実施しうろことが分かる。Now, the core management code is for confirming the validity of the reactor core design and managing the core safely, and since the data mentioned above is based on this management code, its reliability is sufficiently high. If the fuel to be transported to reprocessing facility B is determined based on the concentration of fissile nuclides in the dissolved 1,114 nuclides in reprocessing facility B, the fissile nuclides in melting tank 4 can be The concentration of nuclides depends on the weight of fissile nuclides contained in the fuel, that is, the shear pieces, which are carried out from the storage facility A, loaded into the shear 813, and transferred to the dissolution tank 4. You can see how reliable it is.
今、測定７３４ａにより求められ演算部２ｃに入力され
た現在の溶解槽４内の核種の濃度をＮ′、臨界防止上の
核種の濃度の制限値或はプロセス管理上の目標値、即ち
基準値をＮＬ、次のステップで溶解槽４に移送される剪
断小片群の核種の濃度をＮとすると、Ｎ’＋Ｎ＜ＮＬで
あることが溶解槽４の臨界安全性に必要である。従って
、Ｎ＜ＮＬ−Ｎ’としてＮを求めることができ、メモリ
ー部２ａに記憶された情報よりこのＮに相当する剪断予
定燃料を有する燃ｔ１集合体を泗択する。ＰＡえば、第
２図を参照して、ある使用済み燃料の上部１／４が剪断
されると仮定した場合、メモリー部２ａに記憶された炉
心管理コード出力の核種の軸方向重量分布（ハ）より、
剪断予定燃料＋１＋における核***性核種の重量ＮＩＬ
、ＮＩ２、・・・ＮＩＮが求まり、軸方向燃焼度分布（
ロ）より、剪断すべき燃料ｍ１の燃焼度Ｂｕが求まり、
同様にして、ｍ２、＋１３、論、についてもＢｕ、ｎｕ
、 Ｂｕが得られるので、燃料集合体の平均燃焼度ｆｌ
ｕも次式に従って得られる。Now, the current nuclide concentration in the dissolution tank 4 obtained by the measurement 734a and input to the calculation unit 2c is N', and the nuclide concentration limit value for criticality prevention or the target value for process management, that is, the reference value. Assuming that NL is the nuclide concentration of the sheared particles transferred to the dissolution tank 4 in the next step, N'+N<NL is required for the criticality safety of the dissolution tank 4. Therefore, N can be determined as N<NL-N', and the fuel t1 assembly having the fuel to be sheared corresponding to this N is selected from the information stored in the memory section 2a. PAFor example, referring to FIG. 2, if it is assumed that the upper quarter of a certain spent fuel is sheared, the axial weight distribution of nuclides in the core management code output stored in the memory unit 2a (c) Than,
Weight of fissile nuclides in shear scheduled fuel +1+ NIL
, NI2, ... NIN is determined, and the axial burnup distribution (
From (b), the burnup Bu of the fuel m1 to be sheared is determined,
Similarly, for m2, +13, and Bu, nu
, Bu is obtained, so the average burnup fl of the fuel assembly is
u is also obtained according to the following formula.
これを燃料の決定の指標とすることもできる。ここで、
剪断予定燃料−，よりなる仮定上の剪断小片群中の核種
の濃度Ｎは、ＮＩＩ＋Ｎ１２＋・・・ＮＩＮ ＝Ｎから
求めることができる。このようにして、演算部２ｃにお
いて、再処理施設Ｂへ搬出される燃料を決定したら、メ
モリー部２ａに記憶された当該燃料の燃料番号（イ）及
び燃料配置位！（ホ）に基づいて、燃料移送装ｒＩｔ１
を当該燃料の位置に移動し再処理施設Ｂへ搬出すること
ができる。従って、この燃料集合体を剪断し剪断小片を
溶解槽中に移送しても、溶解槽の臨界安全性や溶解プロ
セスには影響がなく、使用済み燃料貯蔵施設Ａから再処
理施設Ｂの剪断機３への燃料集合体の移送と該剪断Ｒ３
から溶解槽４への燃料剪断片の供給とを一貫して安定的
に行うことができる。This can also be used as an index for determining fuel. here,
The concentration N of nuclides in a hypothetical group of sheared particles consisting of fuel to be sheared can be determined from NII+N12+...NIN=N. In this way, when the calculation unit 2c determines the fuel to be transported to the reprocessing facility B, the fuel number (a) and fuel arrangement position of the fuel are stored in the memory unit 2a. Based on (e), fuel transfer device rIt1
can be moved to the location of the fuel and transported to reprocessing facility B. Therefore, even if this fuel assembly is sheared and the sheared pieces are transferred to the melting tank, there is no effect on the criticality safety of the melting tank or the melting process. Transfer of the fuel assembly to R3 and the shear R3
It is possible to consistently and stably supply the fuel shear pieces from to the melting tank 4.
［発明の効果］
以上のように、本発明によれば、燃焼度を測定すること
なく、炉心管理コードのデータを利用して、再処理施設
における使用済み燃料の核種の濃度から、再処理施設へ
搬出する使用済み燃料を決定するので、再処理施設へ搬
入する際の燃焼度測定にかかる時間をなくし作業の円滑
化を図ると共に、使用済み燃料貯蔵施設から搬出する段
階で再処理施設の臨界安全管理、プロセス管理の作業能
力を高め、再処理施設のプラント稼動率を向上させるこ
とができる。[Effects of the Invention] As described above, according to the present invention, the reprocessing facility can be estimated based on the concentration of nuclides in the spent fuel at the reprocessing facility by using the data of the core management code without measuring the burnup. Since the spent fuel is determined to be transported to the reprocessing facility, it is possible to eliminate the time required to measure the burnup when transporting it to the reprocessing facility, making the work smoother. It is possible to improve the work capacity of safety management and process management, and improve the plant operation rate of reprocessing facilities.
第１図は、本発明の代用済み燃料搬出方法を実施するシ
ステムの概要を示すブロック図、第２図は、剪断される
使用済み燃料を示す説明図である。
Ａ・・・使用済み燃料貯蔵施設
Ｂ・・・再処理施設 １・・・燃料移送装置２・・
・制御ユニット ３・・・剪断機４・・・溶解槽
２ａ・・・メモリー部２ｂ・・・入力部
２ｃ・・・演算部間
（初） （平力燃蚊長）FIG. 1 is a block diagram showing an overview of a system for carrying out the substituted fuel removal method of the present invention, and FIG. 2 is an explanatory diagram showing spent fuel being sheared. A... Spent fuel storage facility B... Reprocessing facility 1... Fuel transfer device 2...
・Control unit 3... Shearing machine 4... Dissolution tank
2a...Memory section 2b...Input section
2c...Between calculation sections (first time) (Heiriki Momocho)
Claims (1)
、その測定値を制御ユニットの演算部に入力し、該演算
部において、前記測定値と前記再処理施設における既知
の核種濃度基準値とから、前記制御ユニットのメモリー
部に記憶された使用済み燃料貯蔵施設内の使用済み燃料
の核種の軸方向重量分布に基づいて、前記使用済み燃料
貯蔵施設から前記再処理施設へ搬出される使用済み燃料
を決定し、該使用済み燃料を再処理施設へ搬出する使用
済み燃料搬出方法。The nuclide concentration of the spent fuel in the reprocessing facility is measured, the measured value is input to the calculation section of the control unit, and the calculation section calculates the nuclide concentration from the measured value and the known nuclide concentration standard value in the reprocessing facility. , the spent fuel to be transported from the spent fuel storage facility to the reprocessing facility based on the axial weight distribution of nuclides in the spent fuel in the spent fuel storage facility stored in the memory section of the control unit. A method for transporting spent fuel by determining the amount of fuel and transporting the spent fuel to a reprocessing facility.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP5873987A JPH0810270B2 (en) | 1987-03-16 | 1987-03-16 | How to carry out spent fuel to the reprocessing facility |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP5873987A JPH0810270B2 (en) | 1987-03-16 | 1987-03-16 | How to carry out spent fuel to the reprocessing facility |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS63225194A true JPS63225194A (en) | 1988-09-20 |
| JPH0810270B2 JPH0810270B2 (en) | 1996-01-31 |
Family
ID=13092886
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP5873987A Expired - Lifetime JPH0810270B2 (en) | 1987-03-16 | 1987-03-16 | How to carry out spent fuel to the reprocessing facility |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH0810270B2 (en) |
Cited By (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2003035795A (en) * | 2001-07-19 | 2003-02-07 | Toshiba Corp | Reprocessing method for reactor fuel, determination method for processing order, processing planning device and program |
| JP2009133701A (en) * | 2007-11-30 | 2009-06-18 | Toshiba Corp | Criticality safety control method for continuous dissolver in reprocessing facility |
| JP2011209144A (en) * | 2010-03-30 | 2011-10-20 | Toshiba Corp | Method, system and program for planning fuel processing |
| JP2012098311A (en) * | 2012-02-24 | 2012-05-24 | Toshiba Corp | Criticality safety management method of continuous dissolver in nuclear reprocessing facilities |
-
1987
- 1987-03-16 JP JP5873987A patent/JPH0810270B2/en not_active Expired - Lifetime
Cited By (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2003035795A (en) * | 2001-07-19 | 2003-02-07 | Toshiba Corp | Reprocessing method for reactor fuel, determination method for processing order, processing planning device and program |
| JP4643066B2 (en) * | 2001-07-19 | 2011-03-02 | 株式会社東芝 | Reactor fuel reprocessing method, processing sequence determination method, fuel processing planning apparatus and program |
| JP2009133701A (en) * | 2007-11-30 | 2009-06-18 | Toshiba Corp | Criticality safety control method for continuous dissolver in reprocessing facility |
| JP2011209144A (en) * | 2010-03-30 | 2011-10-20 | Toshiba Corp | Method, system and program for planning fuel processing |
| JP2012098311A (en) * | 2012-02-24 | 2012-05-24 | Toshiba Corp | Criticality safety management method of continuous dissolver in nuclear reprocessing facilities |
Also Published As
| Publication number | Publication date |
|---|---|
| JPH0810270B2 (en) | 1996-01-31 |
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