JPS6327796A - Criticality controller - Google Patents
Criticality controllerInfo
- Publication number
- JPS6327796A JPS6327796A JP61171189A JP17118986A JPS6327796A JP S6327796 A JPS6327796 A JP S6327796A JP 61171189 A JP61171189 A JP 61171189A JP 17118986 A JP17118986 A JP 17118986A JP S6327796 A JPS6327796 A JP S6327796A
- Authority
- JP
- Japan
- Prior art keywords
- concentration
- input device
- criticality
- solution
- arithmetic processing
- 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
- 238000012545 processing Methods 0.000 claims description 21
- 239000011824 nuclear material Substances 0.000 claims description 20
- 238000000034 method Methods 0.000 claims description 19
- 238000011156 evaluation Methods 0.000 claims description 11
- 238000012958 reprocessing Methods 0.000 claims description 11
- 238000003860 storage Methods 0.000 claims description 8
- 239000002915 spent fuel radioactive waste Substances 0.000 claims description 6
- 239000000126 substance Substances 0.000 claims description 5
- 239000007788 liquid Substances 0.000 description 7
- 238000000605 extraction Methods 0.000 description 5
- 239000002253 acid Substances 0.000 description 4
- 238000013461 design Methods 0.000 description 4
- 238000000638 solvent extraction Methods 0.000 description 3
- WYKYKTKDBLFHCY-UHFFFAOYSA-N chloridazon Chemical compound O=C1C(Cl)=C(N)C=NN1C1=CC=CC=C1 WYKYKTKDBLFHCY-UHFFFAOYSA-N 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 238000012840 feeding operation Methods 0.000 description 2
- 230000004992 fission Effects 0.000 description 2
- 238000012432 intermediate storage Methods 0.000 description 2
- 238000004088 simulation Methods 0.000 description 2
- 239000004215 Carbon black (E152) Substances 0.000 description 1
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 1
- 229910052778 Plutonium Inorganic materials 0.000 description 1
- 229910052770 Uranium Inorganic materials 0.000 description 1
- 239000003153 chemical reaction reagent Substances 0.000 description 1
- 238000005253 cladding Methods 0.000 description 1
- 238000012790 confirmation Methods 0.000 description 1
- 238000005202 decontamination Methods 0.000 description 1
- 230000003588 decontaminative effect Effects 0.000 description 1
- 239000003085 diluting agent Substances 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 239000000446 fuel Substances 0.000 description 1
- 229930195733 hydrocarbon Natural products 0.000 description 1
- 150000002430 hydrocarbons Chemical class 0.000 description 1
- 239000004973 liquid crystal related substance Substances 0.000 description 1
- 239000000463 material Substances 0.000 description 1
- 229910017604 nitric acid Inorganic materials 0.000 description 1
- 238000011017 operating method Methods 0.000 description 1
- 239000003960 organic solvent Substances 0.000 description 1
- 238000005192 partition Methods 0.000 description 1
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 description 1
- 239000002574 poison Substances 0.000 description 1
- 231100000614 poison Toxicity 0.000 description 1
- 238000012552 review Methods 0.000 description 1
- 238000005070 sampling Methods 0.000 description 1
- 239000002904 solvent Substances 0.000 description 1
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 1
- 229910002007 uranyl nitrate Inorganic materials 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
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) [Abstract] This bulletin contains application data before electronic filing, so abstract data is not recorded.
Description
【発明の詳細な説明】
〔発明の目的〕
（産業上の利用分野）
本発明は使用済核燃料の再処理施設における臨界管理装
置に関する。DETAILED DESCRIPTION OF THE INVENTION [Object of the Invention] (Field of Industrial Application) The present invention relates to a criticality control device in a spent nuclear fuel reprocessing facility.
（従来の技術）
使用済核燃料の再処理膿設においては、主手段としての
臨界安全設計と、補助手段としての臨界管理とで臨界事
故を防止している。(Prior art) In spent nuclear fuel reprocessing facilities, criticality safety design is used as the main means and criticality control is used as an auxiliary means to prevent criticality accidents.
このうち、臨界安全設計とは、形状寸法制限、濃度制限
、質量制限および中性子毒の使用、ならびに、これらの
組合わせによって臨界に至らないことを補償することで
あり、例えば、ウランやプルトニウム等の核物質を含む
溶液を受入れるベッセル類については、取扱う核物質の
濃度に拘わらず臨界に達することがないようにする全１
度安全形状寸法に従って設計する。なお、この全濃度安
全形状用法によって設計する方法では装置の寸法が小さ
くなり過ぎて実際の運転に不便になることがある。この
場合には、施設の運転上考えられる種々の核物質濃度を
考慮すると共に、安全上の十分な余裕を見て形状寸法を
制限した設計がなされる。Among these, criticality safety design refers to the use of shape and size restrictions, concentration restrictions, mass restrictions, the use of neutron poison, and the combination of these to ensure that criticality does not occur. Regarding vessels that receive solutions containing nuclear material, all 1 measures must be taken to ensure that the solution does not reach criticality regardless of the concentration of the nuclear material being handled.
Design according to safety dimensions. In addition, in the method of designing using this total concentration safe shape usage method, the dimensions of the device may become too small, making it inconvenient for actual operation. In this case, a design is made that takes into account various nuclear material concentrations that can be considered in the operation of the facility, and limits the shape and size with a sufficient margin for safety.
一方、臨界管理とは、基本的に臨界安全性が確保されて
いる核物質濃度を測定し、ル１１限濃度以下であること
を確認することである。この臨界管理を必要とする例と
して、たとえば、使用済核燃料の再処理法としてのビュ
ーレックス法を採用した再処理施設では、使用済核燃料
を細かく切断して＆０酸で溶解し、燃料被覆材等の不溶
解物を除去した後、酸濃度を調整してパラフィン系炭化
水素希釈剤中の３０％程度のＴＢＰを用いた溶媒抽出に
よりウランとプルトニウムとを核分裂生成物から分離除
去する。この溶媒抽出は核分裂生成物の除染効率を高め
るために酸濃度をうまく調整して抽出、逆抽出、抽出と
いった何段階かの抽出器を通して行なわれるが、逆抽出
後の溶液は一度中間貯槽に入れられて酸濃度の調整が行
なわれる。こうした中間貯槽では、濃度制限による臨界
管理が行なわれることがあり、この臨界管理上は受入れ
液の核物質濃度測定が不可欠である。On the other hand, criticality control basically means measuring the concentration of nuclear material that ensures criticality safety and confirming that it is below the Le 11 limit concentration. For example, in a reprocessing facility that uses the Burex method as a reprocessing method for spent nuclear fuel, spent nuclear fuel is cut into small pieces and dissolved in &0 acid, resulting in fuel cladding and other materials that require this criticality control. After removing insoluble matter, uranium and plutonium are separated and removed from the fission products by solvent extraction using approximately 30% TBP in a paraffinic hydrocarbon diluent while adjusting the acid concentration. In order to increase the decontamination efficiency of fission products, this solvent extraction is carried out through several stages of extraction, extraction, back extraction, and extraction, with the acid concentration well adjusted.The solution after back extraction is once transferred to an intermediate storage tank. The acid concentration is adjusted. In such intermediate storage tanks, criticality control is sometimes carried out by limiting the concentration, and it is essential to measure the nuclear material concentration of the receiving liquid for criticality control.
（発明が解決しようとする問題点）
臨界事故防止の補助手段として、核物″ｆｆ ＆１度を
測定するには、核物質を含む受入れ溶液を事前にサンプ
リングして分析するか、あるいは受入れ途中の十分に未
臨界な最の段階で受入れ液をサンプリングして分析する
等の方法が考えられる。しかし、このいずれの方法にお
いても、受入れ前に運転操作を一時中断してオフライン
での分析結果を待つことになり、いわゆる、運転操作上
の不連続時間が発生する。(Problem to be solved by the invention) In order to measure the nuclear material "ff & 1 degree" as an auxiliary means to prevent criticality accidents, it is necessary to sample and analyze the received solution containing the nuclear material in advance, or to analyze it in advance during the reception. Possible methods include sampling and analyzing the received liquid at the final stage when it is sufficiently subcritical.However, in either of these methods, operation must be temporarily suspended and the results of off-line analysis must be waited for before receiving the liquid. As a result, so-called discontinuous time in driving operations occurs.
かかる、不連続時間の発生は臨界管理に万全を期すため
に止むを得ないものとされていたが、これが再処理ｔｉ
ｈ設の設備利用率を低下させるという問題点があった。The occurrence of such discontinuous time was thought to be unavoidable in order to ensure criticality control, but this
There was a problem in that it lowered the capacity utilization rate of h facilities.
本発明は上記の問題点を解決するためになされたもので
、臨界管理のための運転操作上の不連続時間を最小に留
め得、これによって再処理施設の設備利用率を大幅に向
上させｔｑる臨界管理装置の提供を目的とする。The present invention has been made in order to solve the above problems, and can minimize the discontinuous time in operation for criticality control, thereby greatly improving the capacity utilization rate of the reprocessing facility. The purpose is to provide a criticality control device that
（問題点を解決するための手段）
本発明は、使用済核燃料を再処理する被処理溶液の組成
条件を入力する第１の入力装置と、再処理工程の運転条
件を入力する第２の入力装置と、事前に計算された分配
係数等に関する評価式を記憶させてある記憶装置と、前
記第１の入力装置によって組成条件が入力された被処理
溶液を前記第２の入力装置によって入力された運転条件
で再処理したときの処理済み溶液の核物質濃度の最大値
を、前記記憶装置に記憶させてある評価式を用いると共
に、前記組成条件および運転条件の誤差範囲を考慮して
算出し、且つ、この核物質ｆ：ｉ度の最大値が許容範囲
以内であるか否かを判別する演算処理装置と、この演算
処理装置の処理結果を出力する出力ＨＨとを備えたこと
を特徴とするものである。(Means for Solving the Problems) The present invention includes a first input device for inputting composition conditions of a solution to be treated for reprocessing spent nuclear fuel, and a second input device for inputting operating conditions of the reprocessing process. an apparatus, a storage device in which evaluation formulas related to distribution coefficients etc. calculated in advance are stored, and a solution to be treated whose composition conditions have been inputted by the first input device and which are inputted by the second input device. Calculating the maximum value of the nuclear material concentration of the treated solution when reprocessed under operating conditions using an evaluation formula stored in the storage device and taking into account the error range of the composition conditions and operating conditions, Further, it is characterized by comprising an arithmetic processing device that determines whether the maximum value of the nuclear material f:i degrees is within a permissible range, and an output HH that outputs the processing result of this arithmetic processing device. It is something.
（作 用）
この発明においては、第１の入力装置によって被処理溶
液の核物質濃度等の組成条件を入力する一方、第２の入
力装置によって処理工程の供給液晶、溶媒等の試薬液量
および処理装置の温度、圧力等の運転条件を入力すると
、演算処理装置がこれらの条件および記憶装置に記憶さ
れた評価式を用いて処理工程をシミュレーションして処
理済溶液の核物質濃度の最大値を算出し、且つ、この核
物質濃度の最大値が許容範囲以内であるか否かを判別し
、続いてこれらの処理結果を出力装置によってオペレー
タに知らせる。(Function) In the present invention, while the first input device inputs the composition conditions such as the nuclear substance concentration of the solution to be treated, the second input device inputs the amount of liquid crystal to be supplied in the treatment process, the amount of reagent liquid such as a solvent, etc. When operating conditions such as temperature and pressure of the processing equipment are input, the processing unit simulates the processing process using these conditions and the evaluation formula stored in the storage device, and calculates the maximum value of the nuclear material concentration in the processed solution. It is determined whether the maximum value of the nuclear material concentration is within a permissible range, and then these processing results are notified to the operator by an output device.
しかして、処理溶液を貯留する中間槽の臨界管理をする
場合でも、分析結果を待つという運転操作上の不連続時
間を最小に留めることができる。Therefore, even when performing criticality control of the intermediate tank storing the processing solution, it is possible to minimize the discontinuous time in operation of waiting for analysis results.
（実施例）
第１図は本弁明の一実施例の構成を示すブロック図であ
る。同図において、再処理施設の化学処理工程で処理さ
れる被処理溶液の核物質濃度等の組成条件を入力するだ
めのオペレータコンソール１と、化学処理工程の運転条
件を入カザるプロセス入力装置２と、各団のデータを保
存する記憶装置３とが演算処理装置４に接続され、さら
に、演算処理装置４には処理結架を出力する出力装置５
が１ａ続されている。(Embodiment) FIG. 1 is a block diagram showing the configuration of an embodiment of the present invention. In the figure, there is an operator console 1 for inputting composition conditions such as nuclear material concentration of a solution to be treated in a chemical treatment process of a reprocessing facility, and a process input device 2 for inputting operating conditions for the chemical treatment process. and a storage device 3 for storing data of each group are connected to an arithmetic processing device 4, and an output device 5 for outputting processed data to the arithmetic processing device 4.
is continued 1a.
ここで、記１装置３には事前に計算された分配係数等に
関する評価式も記憶されており、演算処理装置４はこの
評価式を用いて処理工程をシミュレーションするように
なっている。Here, the device 3 also stores evaluation formulas regarding distribution coefficients and the like calculated in advance, and the arithmetic processing device 4 simulates the processing steps using this evaluation formula.
上記のように構成された本実施例の動作を演ｎ処理装置
４の処理手順を示す第２図のフローチャートに従って説
明する。The operation of this embodiment configured as described above will be explained with reference to the flowchart of FIG. 2 showing the processing procedure of the arithmetic processing device 4.
先ず、ステップ１０１にてオペレータがオペレータコン
ソール１によって化学処理工程の被処理溶液の組成条件
を入力すると、ステップ１０２にて演算処理装置４はプ
ロセス入力装置２を介して処理工程の運転条件を入力し
てその監視を続け、ステップ１０３にてこの運転条件の
変動範囲を評価する。なお、組成条件の誤差範囲は通常
一定範囲として記憶されており、その結果をステップ１
０４にて取出す。First, in step 101, the operator inputs the composition conditions of the solution to be treated in the chemical treatment process through the operator console 1. In step 102, the arithmetic processing unit 4 inputs the operating conditions for the treatment process through the process input device 2. Then, in step 103, the range of variation in the operating conditions is evaluated. Note that the error range for composition conditions is usually stored as a fixed range, and the results are used in step 1.
Take it out at 04.
次に、ステップ１０５にて当該処理工程の処理済溶液の
想定最大核物質濃度をｒＰｌｊｌｉする。このとき、ス
テップ１０３．Ｈよび１０４の評！ｉ！ｌｉ値に基づい
てシミュレーションを行うが、このシミュレーションは
記憶装置３に記憶させてある評価式、すなわら、ビュー
レックス法における溶媒抽出工程の有機溶媒と硝酸との
間の硝酸ウラニルの分配係数を用いて近似的に行うこと
ができる。Next, in step 105, the assumed maximum nuclear material concentration of the treated solution in the relevant treatment step is rPljli. At this time, step 103. H and 104 reviews! i! A simulation is performed based on the li value, and this simulation uses the evaluation formula stored in the storage device 3, that is, the partition coefficient of uranyl nitrate between the organic solvent and nitric acid in the solvent extraction step in the Burex method. This can be done approximately using
次に、ステップ１０６では記憶装置３にあらかじめ設定
されている設定許容濃度と評価された想定最大核物質濃
度とを比較し、ステップ１０７にて出力装置５を介して
比較結果を出力する。Next, in step 106, the set allowable concentration preset in the storage device 3 is compared with the estimated maximum nuclear material concentration, and in step 107, the comparison result is outputted via the output device 5.
上述したステップ１０２からステップ１０７までの演算
処理は、実時間に近い形で処理されるため最終の濃度評
価の緒度にはある程度の誤差が含まれる。しかし、その
精度評価の曖昧さは、設定許容濃度に余裕を持たせた形
で補償することにより、臨界管理上の安全性確認の補助
手段として十分に利用できる。Since the arithmetic processing from step 102 to step 107 described above is performed in a manner close to real time, a certain degree of error is included in the final concentration evaluation. However, by compensating for the ambiguity in the accuracy evaluation by leaving a margin in the set allowable concentration, it can be fully utilized as an auxiliary means for safety confirmation in criticality control.
第３図は本実施例を使用して実際の臨界管理を行う場合
の運転操作のフローチャートである。同図において、臨
界管理装置１００により核物質濃度が許容濃度制限値（
１）以下であると判定された場合には、ステップ２０１
で送液操作が行なわれる。一方、核物質濃度が許容濃度
υ１限値（１）以上であればステップ２０２で核物質濃
度がオフライン分析により測定され、ステップ２０３に
て許容濃度制限値（２）以下か否かが判定される。FIG. 3 is a flowchart of operational operations when actual criticality control is performed using this embodiment. In the figure, the nuclear material concentration is determined by the criticality control device 100 to the permissible concentration limit (
1) If it is determined that the following is true, step 201
The liquid feeding operation is performed. On the other hand, if the nuclear material concentration is equal to or higher than the permissible concentration υ1 limit value (1), the nuclear material concentration is measured by off-line analysis in step 202, and it is determined in step 203 whether it is equal to or less than the permissible concentration limit value (2). .
ここで、許容濃度制限値（２）以下と判定された場合に
はじめてステップ２０１の送液操作が行なわれる。一方
、ステップ２０３で核物質濃度が許容濃度ルリ限値（２
）を超えていると判定された場合にはステップ２０４に
て被処理液の希釈操作が行なわれる。Here, the liquid feeding operation in step 201 is performed only when it is determined that the concentration is less than or equal to the allowable concentration limit value (2). On the other hand, in step 203, the nuclear material concentration is determined to be the allowable concentration Lurie limit (2
), the liquid to be treated is diluted in step 204.
なお、この実施例では臨界管理の精度評価の曖昧さを補
償するため制限値（１）を制限値（２）よりも小さく設
定しであるが、通常の運転条件では制限値（１）を超え
る頻度は極めて小さく、従来実施されてきたオフライン
の分析操作を省に’＋ Ｔることもできる。Note that in this example, limit value (1) is set smaller than limit value (2) in order to compensate for ambiguity in criticality control accuracy evaluation, but under normal operating conditions, limit value (1) will be exceeded. The frequency is extremely low, and the off-line analysis operations that have traditionally been performed can be omitted.
以上の説明によって明らかなように、本発明によれば、
運転操作を一時停止させてオフライン分析の結果を待つ
という運転操作上の不連続時間を最小に留め得、これに
よって再処理施設の設備仕様効率を格段に高めることが
できる。As is clear from the above description, according to the present invention,
It is possible to minimize the discontinuous time during which the operation is temporarily stopped and wait for the results of off-line analysis, thereby significantly increasing the equipment specification efficiency of the reprocessing facility.
第１図は本発明の一実施例のｆｆ４成を示すブロック図
、第２図は同実膿例の動作を説明するためのフローチャ
ート、第３図は同実施例の判断結果を用いて臨界管理す
る場合の運転操作手順を示寸フローヂャートである。
１・・・オペレータコンソール、２・・・プロセス入力
装置、３・・・記憶装置、４・・・演算処理装置、５・
・・出力装置。
出願人代理人 佐 藤 −雌
第１図
氾２図Fig. 1 is a block diagram showing the ff4 configuration of an embodiment of the present invention, Fig. 2 is a flow chart for explaining the operation of the same example, and Fig. 3 is a criticality control using the judgment results of the embodiment. This is a flowchart showing the operating procedure when doing so. DESCRIPTION OF SYMBOLS 1... Operator console, 2... Process input device, 3... Storage device, 4... Arithmetic processing unit, 5...
...Output device. Applicant's Representative Sato - Female Figure 1 Figure 2
Claims (1)
する第１の入力装置と、再処理工程の運転条件を入力す
る第２の入力装置と、事前に計算された分配係数等に関
する評価式を記憶させてある記憶装置と、前記第１の入
力装置によって組成条件が入力された被処理溶液を前記
第２の入力装置によって入力された運転条件で再処理し
たときの処理済み溶液の核物質濃度の最大値を、前記記
憶装置に記憶させてある評価式を用いると共に、前記組
成条件および運転条件の誤差範囲を考慮して算出し、且
つ、この核物質濃度の最大値が許容範囲以内であるか否
かを判別する演算処理装置と、この演算処理装置の処理
結果を出力する出力装置とを備えたことを特徴とする臨
界管理装置。A first input device for inputting the composition conditions of a solution to be treated for reprocessing spent nuclear fuel, a second input device for inputting operating conditions for the reprocessing process, and an evaluation formula regarding a pre-calculated distribution coefficient, etc. and a nuclear substance of the treated solution when the solution to be treated whose composition conditions have been inputted by the first input device is reprocessed under the operating conditions inputted by the second input device. The maximum value of the nuclear material concentration is calculated using the evaluation formula stored in the storage device, taking into account the error range of the composition conditions and operating conditions, and the maximum value of the nuclear material concentration is within the permissible range. 1. A criticality management device comprising: an arithmetic processing device that determines whether or not the arithmetic processing device exists; and an output device that outputs a processing result of the arithmetic processing device.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61171189A JPS6327796A (en) | 1986-07-21 | 1986-07-21 | Criticality controller |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61171189A JPS6327796A (en) | 1986-07-21 | 1986-07-21 | Criticality controller |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS6327796A true JPS6327796A (en) | 1988-02-05 |
| JPH0566998B2 JPH0566998B2 (en) | 1993-09-22 |
Family
ID=15918654
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP61171189A Granted JPS6327796A (en) | 1986-07-21 | 1986-07-21 | Criticality controller |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS6327796A (en) |
Cited By (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPH0395493A (en) * | 1989-09-07 | 1991-04-19 | Toshiba Corp | Criticality preventive device |
| EP0631290A1 (en) * | 1993-06-24 | 1994-12-28 | Hitachi, Ltd. | Reprocessing plant and method of operating the same |
| 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 |
| JP2012098311A (en) * | 2012-02-24 | 2012-05-24 | Toshiba Corp | Criticality safety management method of continuous dissolver in nuclear reprocessing facilities |
-
1986
- 1986-07-21 JP JP61171189A patent/JPS6327796A/en active Granted
Cited By (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPH0395493A (en) * | 1989-09-07 | 1991-04-19 | Toshiba Corp | Criticality preventive device |
| JP2809739B2 (en) * | 1989-09-07 | 1998-10-15 | 株式会社東芝 | Criticality prevention device |
| EP0631290A1 (en) * | 1993-06-24 | 1994-12-28 | Hitachi, Ltd. | Reprocessing plant and method of operating the same |
| EP0817207A1 (en) * | 1993-06-24 | 1998-01-07 | Hitachi, Ltd. | Reprocessing plant and method for operating the same |
| 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 |
| 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 |
|---|---|
| JPH0566998B2 (en) | 1993-09-22 |
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Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| EXPY | Cancellation because of completion of term |