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論文

Pipe cutting method at high radiation area in FUGEN

瀧谷 啓晃; 石山 正弘; 手塚 将志; 北山 尚樹

Proceedings of International Conference on Dismantling Challenges; Industrial Reality, Prospects and Feedback Experience (DEM 2018) (Internet), 8 Pages, 2018/10

「ふげん」では、炉心解体に向けた準備及び炉内試料採取のための環境整備の一環として、2015年から2017年にかけて炉心周辺の系統である原子炉冷却系、重水系、ヘリウム系の配管を切断することで炉心の隔離を実施してきた。この隔離作業では、(1)高線量率エリア(空気中で1$$sim$$5mSv/h、配管表面で最大10mSv/h)での作業時間の短縮、(2)重水系及びヘリウム系の内部には20$$sim$$30Bq/cm$$^{3}$$のトリチウムが含まれるため、切断作業中の作業エリアへのトリチウムの拡散防止、(3)原子炉の放射能の評価を考慮して、炉内構造物の汚染状況に与える汚染ヒュームの影響の最小化、といった3つの課題があった。本件は、これらの問題を考慮して、高線量率エリアで重水系及びヘリウム系の配管を切断する方法を検討し、その方法で隔離作業を実施した結果について報告するものである。

論文

The CMMR program; BWR core degradation in the CMMR-3 test

山下 拓哉; 佐藤 一憲; 阿部 雄太; 中桐 俊男; 石見 明洋; 永江 勇二

Proceedings of International Conference on Dismantling Challenges; Industrial Reality, Prospects and Feedback Experience (DEM 2018) (Internet), 11 Pages, 2018/10

2011年に発生した福島第一原子力発電所事故における、燃料集合体の溶融進展挙動については、未だ十分に解明されていない。1979年に発生したスリーマイル島原子力発電所2号炉の事故以降、加圧水型原子炉を中心としたシビアアクシデントについては、炉心溶融の初期挙動や圧力容器破損に関わる個別現象に着目した試験が多数行われてきた。しかし、炉心溶融が進行し、炉心物質が炉心から下部プレナムへと移行する過程に関わる既往研究は少なく、特に、この移行経路に制御棒と複雑な炉心下部支持構造が存在する沸騰水型原子炉(以下、「BWR」という)条件での試験データはほとんどない。本研究では、UO$$_{2}$$ペレットの代りにZrO$$_{2}$$ペレットを用いた燃料集合体規模の試験体に対し、BWR実機で想定される軸方向温度勾配をプラズマ加熱により実現し、高温化炉心のガス透過性および高温化炉心物質の支持構造部への進入と加熱を明らかにするための試験を実施した。その結果、高温化した炉心燃料は、部分的な閉塞を形成するが、残留燃料柱は互いに融着しない傾向が強く、崩壊した場合を含めて気相(及び液相)に対するマクロな透過性を持つことが明らかとなった。

口頭

Development of treatment method for analytical waste solutions in STRAD project, 1; Oxidative decomposition of ammonium ion with catalyst

粟飯原 はるか; 渡部 創; 野村 和則; 神谷 裕一*

no journal, , 

An effective decomposition method of ammonium ion in radioactive analytical waste solutions containing U, Pu and nitric acid has been developed to suppress formation of ammonium nitrate. Oxidative decomposition was examined using ozone gas with existence of cobalt ions as homogeneous catalyst. Reduction in concentration of ammonium ion was achieved by combination of the ozone gas oxidation with the catalyst, and the behavior distinctly depend on compositions of the initial solution.

口頭

Development of treatment method for analytical waste solutions in STRAD project, 3; Separation of ammonia by vaporization

松浦 治明*; 小林 亜美*; 三善 真秀*; 粟飯原 はるか; 渡部 創; 野村 和則

no journal, , 

Treatment of radioactive waste solution containing various chemical reagents would be critical issue for decommissioning of nuclear facilities. We have targeted on the wastes containing ammonium ion which would produce ammonium nitrate. To decompose ammonium ion by catalysis, separation of ammonium ion in the solution is expected to be achieved. By using simulated waste solutions, vaporization behavior of ammonia has been investigated. As a result, about 40% of ammonia in the simulated solution with NH$$_{4}$$$$^{+}$$ of 4.47 g/L was recovered by adjustment in pH and heating.

口頭

Development of treatment method for analytical waste solutions in STRAD project, 2; Ammonium ion adsorption onto zeolites

浅沼 徳子*; 宮野 陸*; 粟飯原 はるか; 渡部 創; 野村 和則

no journal, , 

In order to avoid production of ammonium nitrate, stabilization of ammonium ion in analytical waste solution is necessary. In this study, adsorption separation of ammonium ion from the solution was made by using zeolites, i.e. Clinoptilolite, IE-96, HiSiv$$^{TM}$$1000. Fundamental ad-sorption characteristics of these zeolites which is adsorption rate, isotherm analysis and pH property were investigated.

口頭

Radioactive nuclides recovery from spent solvent in STRAD project, 1; Zr recovered by alkaline solutions

荒井 陽一; 渡部 創; 久保田 俊夫*; 野村 和則

no journal, , 

In order to recover U and Pu from the degraded solvents as a part of waste liquid management program STRAD (Systematic Treatments of RAdioactive liquid wastes for Decommissioning), recovery of loaded cations in simulated degraded solvents into alkali solutions were experimentally examined. A target solvent of this study was TRUEX solvent involving CMPO and TBP in n-dodecane. Zr was used as a simulant of Pu, and recovery performance and mechanism of Zr from simulated spent solvent into NaOH or Na$$_{2}$$CO$$_{3}$$ solutions were evaluated. Zr was recovered in NaOH solution as precipitate, while back extraction of Zr into Na$$_{2}$$CO$$_{3}$$ solution was progressed by forming carbonate complex. Similar recovery mechanisms are also expected for U and Pu loaded in the solvent.

口頭

Radioactive nuclides recovery from spent solvent in STRAD project, 2; Zr adsorption onto ion exchange resins

中村 文也*; 安倍 諒治*; 新井 剛*; 瀬古 典明*; 荒井 陽一; 渡部 創; 野村 和則

no journal, , 

It is considered that tributyl phosphate (TBP) contained in spent PUREX solvent is well known to decompose into dibutyl phosphate (DBP) by radiation exposure. It has been found that DBP extracts U and Pu to form strong complexes. Since these complexes have stability even at low nitric acid solution, the spent solvent loaded with U and Pu for a long term is difficult to back extracted using nitric solution. Therefore, we have studied the treatment method for adsorbing of U and Pu from spent solvent using adsorbent. In this study, the method of Zr(IV) quantitative analysis using fluorescence spectroscopy was established. Moreover, adsorption behavior of Zr(IV) in DSP-n using adsorbents was evaluated. It was confirmed that non-woven fabrics with phosphoric acid type and imino diacetic acid type showed excellent adsorption ability.

口頭

Overview of STRAD project for systematic treatments of radioactive liquid wastes generated in nuclear facilities

渡部 創; 小木 浩通*; 荒井 陽一; 粟飯原 はるか; 柴田 淳広; 野村 和則; 神谷 裕一*; 浅沼 徳子*; 松浦 治明*; 久保田 俊夫*; et al.

no journal, , 

A new collaborative research project for systematic treatments of radioactive liquid wastes containing various reagents generating in nuclear facilities was started from 2018 initiated by Japan Atomic Energy Agency. The project was named as STRAD (Systematic Treatments of RAdioactive liquid wastes for Decommissioning) project. Tentative targets to be studied under the project are aqueous and organic liquid wastes which have been generated by experiments and analyses in a reprocessing experimental laboratory of JAEA. Currently fundamental studies for treatments of the liquid wastes with complicated compositions are underway.

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