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阿部 拓海; 西原 健司
Progress in Nuclear Science and Technology (Internet), 8, p.47 - 51, 2025/09
The robustness of an entire nuclear fuel cycle (NFC) can be assessed through simulations of the operational factors (OFs) of future NFC facilities, combined with mass flow analyses assuming many time series of OFs. In this study, the uncertainty of OF caused by minor troubles, which causes the expansion of the regular maintenance or temporary suspension, was focused on. OF of a reprocessing plant with the uncertainty were predicted by autoregressive moving average model. As a demonstration of the methodology to assess the robustness of an NFC, using the predicted OF data and a NFC simulator, NMB code, the impact of a reprocessing plant OF on a fast reactor OF was quantified. As a results, extra reprocessing capacity or additional plutonium stock induced higher robustness of an NFC.
-hexaoctyl nitrilotriacetamide (HONTA) using mixer-settler extractors伴 康俊; 鈴木 英哉*; 宝徳 忍; 津幡 靖宏
Progress in Nuclear Science and Technology (Internet), 8, p.243 - 247, 2025/09
A continuous counter-current extraction experiment was performed by mixer-settler extractors to recover minor actinides (MA; Am and Cm) from high-level liquid waste. Using -hexaoctyl nitrilotriacetamide (HONTA) as an extractant, 0.17 g of MA was recovered in a MA fraction.
佐藤 理花; 近藤 俊樹; 梅田 良太; 菊地 晋; 山野 秀将
Progress in Nuclear Science and Technology (Internet), 8, p.137 - 142, 2025/09
ナトリウム-溶融塩熱交換器を有する蓄熱式高速炉では、ナトリウム(Na)と硝酸系溶融塩との熱交換器伝熱管破損に至るような仮想的な事故条件下でNaと硝酸系溶融塩との化学反応が発生する可能性がある。そのため、Naと硝酸系溶融塩の反応挙動は、当該システムの安全評価上、重要現象の一つとなっている。本研究では、NaNO
-KNOの混合物であるソーラーソルトとNaとの反応試験を実施し、得られた試験結果について検討を行った。その結果、ソーラーソルトの融解が開始した後にNaとの反応が起こることが分かった。試験で得られた反応温度から、速度論的パラメータおよび反応速度を求め、Na-水反応と比較した。その結果、Na-溶融塩熱交換器を有する蓄熱式高速炉の伝熱管破損時の事象進展で勘案すべき時間スケール内にソーラーソルト反応が生じ得ることが分かった。
山本 昌彦; 堀籠 和志; 後藤 雄一; 田口 茂郎; 久野 剛彦
Progress in Nuclear Science and Technology (Internet), 8, p.387 - 392, 2025/09
東海再処理施設(TRP)では本格的な廃止措置に向けて工程洗浄を実施し、2024年2月に完了した。TRPの主要工程には核物質が残存しているため、高放射性廃液貯槽への移送及びウラン溶液の三酸化ウランへの転換により核物質を回収し、硝酸及び純水による関連工程を洗浄した。この作業では、核物質管理の計量管理のため、工程洗浄の状況に応じてウラン、プルトニウムを同位体希釈質量分析法、重量分析法、分光光度法、アルファ線計数法などの分析法で実施した。また、将来の系統除染に備え、洗浄液中のガンマ線放出核種を高純度ゲルマニウム検出器で測定した。本報告書は、これら再処理施設の工程洗浄にかかわる分析及びその結果をまとめたものである。
竹下 健二*; 岡村 知拓*; 中瀬 正彦*; 阿部 拓海; 西原 健司
Progress in Nuclear Science and Technology (Internet), 8, p.52 - 57, 2025/09
This study analyzed minor actinide (MA) inventory in scenarios assuming demonstration and subsequent commercialization of fast reactor (FR) in the mid-21st century, focusing on the characteristics of reprocessing processes in oxide and metal fuel FR cycles. At the end of the evaluation period defined in this study, the transition of MA to waste was 138 tons in the oxide fuel FR cycle without an MA separation process, requiring a footprint of geological repository of 3.01 km
. In contrast, in the metal fuel FR cycle, when only spent fuel discharged from the FR was subjected to pyro-reprocessing, the MA transition to waste was nearly identical to that of the oxide FR cycle. However, when spent MOX fuel discharged from light water reactor (LWR) was also reduced to metal and processed by the pyro reprocessing, the MA transition decreased to 93 tons, with a correspondingly reduced footprint of 2.12 km2. The results show a strong link between MA transition to waste and repository footprint, highlighting the potential of metal fuel FR cycles which can reduce demand of final disposal by the metallization and pyro-reprocessing of spent MOX fuel from the LWR fuel cycle.
中原 将海; 先崎 達也; 佐野 雄一; 加藤 正人
Progress in Nuclear Science and Technology (Internet), 8, p.64 - 69, 2025/09
高速炉燃料サイクルではマイナーアクチニドを回収し、燃料として再利用することを提案している。本研究では抽出クロマトグラフィ法において高レベル放射性廃液から回収したマイナーアクチニド溶液とウラン及プルトニウム硝酸溶液を混合し、マイクロ波加熱によりマイナーアクチニド含有混合酸化物燃料粉末を調製した。また、X線回折と熱分析によりその特性を評価した。
粟飯原 はるか; 比内 浩; 柴田 淳広; 富田 さゆり*; 駒 義和
Progress in Nuclear Science and Technology (Internet), 8, p.324 - 328, 2025/09
福島第一原子力発電所に滞留数する汚染水にはPuとAmが含まれ、建屋コンクリートを汚染している。汚染の状態を推定するために、汚染のメカニズムを調べることは非常に重要である。そのためセメントペーストと骨材のへのPuとAmの分配比を実験的に求めた。セメントペーストと骨材をPuとAmが含まれる溶液に浸漬し、分配比を取得した。PuとAmのセメントペーストへの分配比は高い値を示し、建屋コンクリートに収着して蓄積していることが示唆された。
Bachmann, A. M.*; 西原 健司; Richards, S.*; 阿部 拓海; Feng, B.*
Progress in Nuclear Science and Technology (Internet), 8, p.11 - 16, 2025/09
Verification exercises between fuel cycle simulators are important for understanding how the methodology and capability differences between the simulators affect the results. This work performs an initial verification exercise with the Cyclus and NMB fuel cycle simulators. The exercise compares the results of the two codes in three simple fuel cycle scenarios: a once-through scenario with a pressurized water reactor, a limited recycle scenario with a pressurized water reactor, and a continuous recycle scenario with a pressurized water reactor and a sodium fast reactor. The results of this exercise highlight the differences in the codes' methodologies to determine when fresh fuel is fabricated and to model fuel depletion, affecting where material is located in a scenario.
佐藤 淳也; 高橋 裕太; 砂原 淳*; 齋藤 利充*; 吉田 幸彦; 曽根 智之; 大杉 武史
Progress in Nuclear Science and Technology (Internet), 8, p.307 - 312, 2025/09
For low-level radioactive waste containing hazardous substances (mixed wastes) generated from uranium fuel processing plants, establishing appropriate solidification/stabilization methods is one of the key challenges for their safe and effective storage and disposal in Japan. This study investigated the solidification/stabilization methods of the mixed waste sludge containing hazardous substances of lead, cadmium and mercury by using various solidification materials. Additionally, the compressive strength of solidified products was investigated along with the leaching behavior of hazardous substances.
岩佐 龍磨; 高野 公秀
Progress in Nuclear Science and Technology (Internet), 8, p.291 - 295, 2025/09
日本原子力研究開発機構では、分散型燃料に添加するMA窒化物粒子の作製技術として、外部ゲル化法による粒子作製技術の開発を実施している。粒子作製技術として外部ゲル化法は一般的に研究されている手法ではあるが、MA窒化物燃料製造に関しては未だデータがほとんど存在しておらず、データ取得が必要である。粒子分散型窒化物燃料について研究した過去の報告書より、燃料の熱物性の低下を避けるためには、添加粒子のサイズは直径250マイクロメーターよりも小径であることが望ましいとされており、本研究においては、後の焙焼及び窒化工程で粒子径が半分以下に縮むことを考慮した上で、500マイクロメーターより小さな球状ゲル粒子を作製可能な、外部ゲル化法における最適条件について検討した。試験条件としては、試験溶液の粘度及び滴下圧力をパラメータとして様々に変化させた。結果として、溶液の粘度を30cPから50cpまで変化させた際に、それぞれ相関して350kPaから500kPa以上の圧力においてより小径かつ真球に近いゲル球が得られた。
樋川 智洋; 津幡 靖宏; 熊谷 友多; 伴 康俊
Progress in Nuclear Science and Technology (Internet), 8, p.286 - 290, 2025/09
本発表では、放射線影響を反映した指標を用いた簡便な分離プロセスシミュレーション法を提案する。放射線分解による抽出能劣化を考慮してマイナーアクチノイド分離プロセスのシミュレーションを行った。その結果、プロセスでのマイナーアクチノイドの処理限界量が求められ、プロセスの放射線耐性についての知見が得られた。
長谷川 健太; 安倍 弘; 高畠 容子; 渡部 創; 中村 雅弘; 佐野 雄一; 竹内 正行
Progress in Nuclear Science and Technology (Internet), 8, p.248 - 251, 2025/09
Japan Atomic Energy Agency has been working on development of extraction chromatography technology for recovery of trivalent minor actinides (MA(III): Am, Cm) from high-level radioactive liquid waste generated in reprocessing of spent nuclear fuel. In this project, a diaphragm pump with radiation resistant is being developed for use in feeding the liquid on the recovery system. In this study, the degradation behavior of ethylene-propylene-diene (EPDM) rubber, selected as a candidate diaphragm material for diaphragm pumps, was quantitatively evaluated by irradiation tests. The rubber samples were immersed in nitric acid solution under tensile load, and irradiated with gamma rays. After irradiation tests, tensile testing and dynamic mechanical analysis, and so on, were conducted to the rubber samples.
中村 聡志; 鈴木 英哉*; 伴 康俊; 大橋 朗*
Progress in Nuclear Science and Technology (Internet), 8, p.228 - 232, 2025/09
日本原子力研究開発機構では、高レベル放射性廃棄物の減容・有害度低減に向け、高レベル放射性廃液からマイナーアクチノイド(MA)を回収するための分離技術の開発を行っている。分離プロセスにおいて、化学的性質の類似する希土類元素(RE)とAmやCmのようなMA(III)の相互分離は非常に難しい。本研究では、単段バッチ法により3種類のグリシン系アミド酸抽出剤のRE(III)及びMA(III)に対する抽出特性を検討した。すべての金属イオンの分配比は平衡pHの上昇とともに増加し、REよりもAmの方が高い分配比を示す結果が得られた。
高野 和也; 栗坂 健一; 山野 秀将
Progress in Nuclear Science and Technology (Internet), 8, p.82 - 85, 2025/09
ナトリウム-溶融塩熱交換器を有する蓄熱式高速炉のリスク評価技術開発の一環として、溶融塩を活用した既存の太陽熱蓄熱発電システムにおける事故トラブル事例結果に基づき、熱交換器における伝熱管破損件数と溶融塩暴露時間を整理するとともに、ベイズ推定手法に基づき伝熱管破損発生率を評価する方法を検討した。
神野 郁夫; 奥村 啓介; 松村 太伊知; Riyana, E. S.; 寺島 顕一; 坂本 雅洋
Progress in Nuclear Science and Technology (Internet), 8, p.343 - 346, 2025/09
福島第一原子力発電所2号機のシールドプラグ隙間のCs-137汚染分布をオペレーションフロアから測定するために、ガンマ線ピンホールカメラを用いる方法を提案する。決定論的計算により、可能性を議論する。
佐々木 祐二; 金子 政志; 熊谷 友多; 伴 康俊
Progress in Nuclear Science and Technology (Internet), 8, p.202 - 204, 2025/09
2種の抽出剤(TODGA, ADAAM)と1種のマスキング剤(DTBA)が原子力機構で開発された。TODGAはアクチノイド(An)とランタノイド(Ln)の同時抽出、DTBAはAnとLnの相互分離、ADAAMは高いAm/Cm分離比(6)を示す。これらの試薬を使って、LnからAnの有効な分離法やAmの単離を検討した。ここではTODGA, DTBA, ADAAMを使った基礎的な抽出挙動を示し、An+Ln一括抽出、An/Ln分離、Am/Cm分離の有効な水相、有機相条件を提案する。
荒井 陽一; 後藤 泰裕; 渡部 創; 吾郷 友宏*; 新井 剛*; 勝木 健太*; 福元 博基*; 保科 宏行*; 瀬古 典明*
Progress in Nuclear Science and Technology (Internet), 8, p.329 - 332, 2025/09
Radioactive liquid waste containing nuclear fuel materials and chemical reagents is stored in nuclear facilities. To eliminate the radioactivity of the radioactive liquid waste, we developed RFIDA, a new perfluoroalkyl (RF)-based ligand with a basic structure of iminodiacetic acid (IDA). In this study, an adsorption test was conducted by impregnating RFIDA into porous silica with a polymer was conducted to confirm that the synthesized RFIDA adsorbs cations. The results confirmed that RFIDA exhibits the ability to adsorb or elute uranium depending on the nitric acid concentration.
木村 祥紀; 山口 知輝
Progress in Nuclear Science and Technology (Internet), 7, p.60 - 66, 2025/03
Rapid and precise detection or identification of artificial radioisotopes is one of the challenging issues in the field of nuclear detection and nuclear security. Although many handheld instruments capable of automated radioisotope detection and identification using small gamma-ray detectors have been recently employed in the field of nuclear security, the performance of such instruments is suffered from the limitation on cost of their detectors in many cases due to limited efficiency or limited energy resolution of the small size detectors. In this paper, a mobile radiation measurement system using multiple small gamma-ray detectors for radioisotope detection and identification has been proposed. The performance of the proposed mobile system for radioisotope identification has been tested for measured gamma-ray spectra of artificial radioisotopes and nuclear materials by CZT detector and CsI(Tl) detector. By analyzing the combined spectrum to be output by the proposed system, significant improvements in the minimum detection time for peaks from radioisotopes and lower detection limits for nuclear materials have been confirmed, comparing to the spectra measured by individual detectors. The authors have also discussed the optimization of detector combinations in terms of detector cost and improvement of sensitivity performance by performing simulations for several types of gamma-ray detectors.
粟飯原 はるか; 渡部 創; 北脇 慎一; 神谷 裕一*
Progress in Nuclear Science and Technology (Internet), 7, p.175 - 181, 2025/03
Many kinds of radioactive liquid wastes have been generated and accumulated in nuclear research facilities, and disposal or treatment procedures have not been established for those wastes due to their complicated compositions. Japan Atomic Energy Agency (JAEA) is developing technologies for the treatment of radioactive liquid waste containing various chemical species through joint research project with several organizations named as STRAD (Systematic Treatment of RAdioactive liquid wastes for Decommissioning). Chemical processing facility (CPF) in JAEA was selected as a model case, and fundamental studies have been conducted for treatments of the liquid wastes accumulated in the CPF. Decomposition of the ammonium ions (NH
) in advance with formation of explosive ammonium nitrate is one of promising treatments of the analytical liquid wastes. We have been focusing on catalytic ozonation of NH as a suitable method for the decomposition under mild conditions with small secondary waste. Detail reaction mechanism has not been revealed yet. In this study, concentration of each component was parametrically changed and the reaction mechanism was investigated. Test solutions containing NH, Co
and Cl
with various concentrations were prepared, and the ozone oxidation experiments were conducted at 333 K. The oxidation reaction of NH proceeds even under low Cl concentration, it is suggested that once chloramine are formed, thus the chlorine was released by decomposition. The above catalytic ozonation was applied to the genuine radioactive liquid waste stored at CPF, and concentration of NH in the waste solutions was successfully reduced. This experience would contribute to treatment of radioactive liquid wastes in other facilities.
by solvent extraction as a potential method to determine Fe in glass containing Sn菅野 直樹*; 中瀬 正彦*; 西條 佳孝*; 松村 大樹; 辻 卓也; 竹下 健二*; 塚原 剛彦*
Progress in Nuclear Science and Technology (Internet), 7, p.154 - 160, 2025/03
The amounts of Fe and Fe
present in glass are important indicators of its optical properties because even small amounts have significant effects. However, it is challenging to use wet chemical analysis to determine the concentration and the ratio of Fe and Fe in glass when it contains Sn because of the redox reaction between Sn and Fe in the glass decomposition solution. A two-step approach was tested to determine the concentrations of Fe and Fe in a glass decomposition solution in the presence of Sn. In the first step, the redox reaction between Sn and Fe was suppressed by increased pH. In the second step, Sn was removed from the glass decomposition solution by solvent extraction. To understand the kinetics of the redox reaction between Sn and Fe, time-resolved dispersive X-ray absorption fine structure and ultraviolet-visible absorption spectroscopy (UV-vis) were used with standard chloride solutions of Sn and Fe in respective valences. We found that lowering the acid concentration suppressed redox reactions. The partitioning behaviors of Sn, Sn
, Fe, and Fe by bis(2-ethylhexyl) hydrogen phosphate (D2EHPA) as an extractant in 
-dodecane were investigated to see the removal ability of Sn from the glass decomposition solution. As a result, D2EHPA in -dodecane could extract Sn, Sn, and Fe into the organic phase, and Fe remained in the aqueous phase. The simultaneous removal of Sn and Sn and the separation of Fe and Fe became possible. This method can potentially apply to the rapid analysis of the concentration of Fe and Fe in a glass containing tin oxide.
