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山野 秀将; 江村 優軌; 高井 俊秀; 久保 重信; Quaini, A.*; Fossati, P.*; Delacroix, J.*; Journeau, C.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
本論文では、炉心混合物質の相互作用に関する反応速度論、二酸化ウラン(UO
)と鉄(Fe)と単価ホウ素(B
C)に対する高温熱物性データ、BCとステンレス鋼(SS)並びにBC-SS共晶物再配置(固化も)に関する実験研究、シビアアクシデントコードにBC-SS共晶物と反応モデルを組み込む。
Johnson, M.*; 江村 優軌; Clavier, R.*; 松場 賢一; 神山 健司; Brayer, C.*; Journeau, C.*
Nuclear Engineering and Design, 423, p.113165_1 - 113165_14, 2024/07
被引用回数：1 パーセンタイル：82.00(Nuclear Science & Technology)日本原子力研究開発機構のMELT施設において、ナトリウム冷却高速炉のシビアアクシデントに関連する溶融ジェットとナトリウムの相互作用に関する実験的研究を行っている。X線イメージングと固化物の分析により、溶融ジェットと冷却材の接触界面でクラストが急速に形成され、熱的な微粒化現象が誘発されることを明らかにした。溶融ジェットと冷却材の接触界面における熱伝達計算の結果は、冷却材との接触から数ミリ秒以内に固体クラストが形成されることを示唆している。X線イメージングを用いたジェットへの冷却材巻き込みの観察結果に基づき、熱的な微粒化が促進されるメカニズムを提案する。
坪田 陽一; Porcheron, E.*; Journeau, C.*; Delacroix, J.*; Suteau, C.*; Lallot, Y.*; Bouland, A.*; Roulet, D.*; 三次 岳志
Proceedings of International Conference on Environmental Remediation and Radioactive Waste Management (ICEM2023) (Internet), 6 Pages, 2023/10
福島第一原子力発電所(1F)における燃料デブリ取出しを安全に実施するためには、燃料デブリの切断時に発生する放射性微粒子の定量評価が必要である。我々はウラン含有のIn/Ex-Vessel組成を持つ模擬燃料デブリを作製し、それらを加熱した際に生じるエアロゾルの物理、化学的特性を評価した。それらを基に1F-Unit2の燃料デブリを加熱法の代表例であるレーザー切断した際に生じるエアロゾルの同位体組成と放射能を推定したところ、
Pu、
Am、
Cmを主とするプルトニウムが
核種として、Pu、
Cs-Ba、
Sr-Yが
核種としての着目核種であることが分かった。
Porcheron, E.*; Journeau, C.*; Delacroix, J.*; Berlemont, R.*; Bouland, A.*; Lallot, Y.*; 坪田 陽一; 池田 篤史; 三次 岳志
Proceedings of International Conference on Environmental Remediation and Radioactive Waste Management (ICEM2023) (Internet), 5 Pages, 2023/10
福島第一原子力発電所(1F)の損傷した原子炉の廃止措置における重要な課題である燃料デブリの切断作業における放射性エアロゾルの発生と飛散を評価する目的で行われたURASOLプロジェクトにおいて、模擬燃料デブリの加熱および機械的切断に伴い生成するエアロゾルの質量濃度、リアルタイム数密度、質量基準の粒径分布、形態、および化学的特性の観点からの特性評価について報告する。加熱試験においては温度上昇に伴う粒径増大が観察され、粒子数密度に関しては劣化ウランを用いた模擬燃料デブリを用いた例がHf含有模擬燃料デブリを用いた試験よりも小さい数密度であった。機械的切断においてはエアロゾルの空気動力学的質量中央径は、放射性試料と非放射性試料でほぼ同程度(約3.7
4.4
m)であった。
Journeau, C.*; Delacroix, J.*; Gu
var, C.*; Testud, V.*; Brackx, E.*; Porcheron, E.*; Bouland, A.*; Berlemont, R.*; 池田 篤史
Science Talks (Internet), 6, p.100215_1 - 100215_9, 2023/05
One of the important challenges for the decommissioning of the damaged reactors of the Fukushima Daiichi Nuclear Power Station (1F) is the fuel debris retrieval. The URASOL (URAnium and aeroSOL) project has been undertaken by the French consortium laboratories consisting of ONET, CEA, and IRSN for JAEA. It aims at acquiring basic data on the generation and characteristics of radioactive aerosols from the thermal or mechanical processing of fuel debris simulant. Prototypic fuel debris samples were fabricated based on the average of the lower head compositions computed in the OECD/BSAF benchmark. Samples were heated in an induction furnace to simulate thermal cutting and released aerosols were collected during three temperature ramps using impactors. The collected aerosols were chemically analyzed by ICP-AES and ICP-MS. Iron and tin are the major elements found in these aerosols, followed by chromium and silicon. Significant releases of tellurium, barium and cerium were observed.
Journeau, C.*; Molina, D.*; Brackx, E.*; Berlemont, R.*; 坪田 陽一
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 5 Pages, 2022/10
CEAは、劣化ウランを使用したUOまたはHfOを(核燃料としての)UOの 代替として用いて、福島第一原子力発電所の模擬燃料デブリを製造した。溶融燃料-コンクリート相互作用によって生じたEx-vessel模擬燃料デブリでは、酸化物相の密度が金属相の密度より軽くなる。それゆえ重い金属質の相が底に偏析する。このうち3つの金属質試料を、CEAカダラッシュ研究所でのハンドソー切断、及び同研究所のFUJISAN施設でコアボーリング装置により機械的に切断された。これらの金属ブロックのうち、2つは非常に切断しにくく(1つはUO試料、もう1つはHfO試料)、最後の1つはより簡単に切断可能であった。これらの3つの金属ブロックの金相分析(SEM-EDSとXRD)の類似点/相違点に関して議論する予定である。この経験は、福島第一原子力発電所の燃料デブリの切断・回収を視野に入れた場合、有益な学びとなる。
Porcheron, E.*; Leblois, Y.*; Journeau, C.*; Delacroix, J.*; Molina, D.*; Suteau, C.*; Berlemont, R.*; Bouland, A.*; Lallot, Y.*; Roulet, D.*; et al.
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 5 Pages, 2022/10
福島第一原子力発電所(1F)の事故炉廃止措置における重要な課題の一つが、燃料デブリの取り出しである。ONET Technologies, CEA, IRSNからなるフランスのコンソーシアムがJAEA/CLADSのために実施したURASOLプロジェクトは、燃料デブリ模擬物質の熱的・機械的加工による放射性エアロゾルの生成と特性に関する科学的基礎データの取得に取り組んでいる。VITAE施設で行われる加熱試験はレーザーによる熱的切断の代表的な条件を模擬している。機械的切断では、FUJISAN施設においてコアボーリング試験を実施した。燃料デブリ模擬物質は、非放射性試験と放射性試験のために開発されている。化学的特性評価と粒径情報の取得は、デブリ取り出しで発生する可能性のある放射性粒子の特性推定のために実施された。これらの情報は1Fにおける燃料デブリ取り出し作業において放射線防護上の対策を評価するうえで重要な情報である。
久保 重信; Payot, F.*; 山野 秀将; Bertrand, F.*; Bachrata, A.*; Saas, L.*; Journeau, C.*; Gosse, S.*; Quaini, A.*; 柴田 明裕*; et al.
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04
The paper presents major outcomes of the France-Japan ASTRID collaboration and the work program from 2020 to 2024 in the field of severe accident study. In the ASTRID collaboration, severe accident sequences were summarized based on the various safety analyses for the important accident phases, which contributed to strengthen the confidence in the ASTRID severe accident progression for a robust safety demonstration and identification of R and D programs of common interest. Collaborative analysis has been conducted to evaluate ASTRID mitigation device efficiency for mitigation of power excursions, material relocation, debris bed and molten pool behavior on the core catcher. The methodology for mechanical consequence assessment was also developed. In order to support the reactor studies, experimental studies have been planned and conducted regarding the reaction of core material mixtures, in-pile experiments for the fuel pin failure and material relocation through a steel duct structure, and out-of-pile experiments for the fuel coolant interaction (FCI) in the sodium pool. Severe accident analysis tool SIMMER-V with new simulation capabilities and SEASON platform have also been developed. Based on these successful achievements, several tasks to study the large fields of the severe accident domain, which include development of severe accident analysis methodologies and synthesis of SA sequences and consequences, thermodynamics, kinetic and thermo-physical studies of core material mixture, development and validation of SIMMER-V, experimental programs on the molten core material relocation and FCI. After defining the technical contents and implementation plans, the five years study programs have been started.
Johnson, M.*; Delacroix, J.*; Journeau, C.*; Brayer, C.*; Clavier, R.*; Montazel, A.*; Pluyette, E.*; 松場 賢一; 江村 優軌; 神山 健司
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04
シビアアクシデントに関する日仏共同実験の一環として、ナトリウム冷却高速炉の原子炉容器内下部プレナムへ溶融燃料が流出した時の燃料-冷却材相互作用について、その解明に向けた研究を実施している。MELT施設では、ナトリウム中へ流出したキログラム単位の模擬溶融炉心物質が急冷される様子をX線で可視化することができる。現在準備中のSERUA施設では、融体と冷却材の接触境界面温度が上昇した場合の沸騰熱伝達を評価するためのデータ取得を予定している。この論文では、これらの施設を活用した実験協力の現状について紹介する。
Grambow, B.; 二田 郁子; 柴田 淳広; 駒 義和; 宇都宮 聡*; 高見 龍*; 笛田 和希*; 大貫 敏彦*; Jegou, C.*; Laffolley, H.*; et al.
Journal of Nuclear Science and Technology, 59(1), p.1 - 24, 2022/01
被引用回数：24 パーセンタイル：75.68(Nuclear Science & Technology)Following the NPP accident, some hundred tons of nuclear fuel elements of 3 damaged nuclear reactor units were partly molten with even larger masses of steel and concrete structures, creating a big mass of corium and fuels debris. Since ten years, this heat generating mass has been cooled permanently by millions of m
of water flowing over them. Knowledge on the interaction of this solid mass with water is crucial for any decommissioning planning. Starting from analyses of the evolutions of the accident in the 3 reactor cores and associated fuel debris formations and some additional isotopic and physiochemical information of debris fragments collected in soils of Fukushima, we review the temporal evolution of the chemistry and leached radionuclide contents of the cooling water. Measured concentration ratios of the actinides and fission products in the water where compared to reported results of laboratory leaching studies with either spent nuclear fuel or simulated fuel debris under a variety of simulated environmental conditions.
Johnson, M.*; Journeau, C.*; 松場 賢一; 江村 優軌; 神山 健司
Annals of Nuclear Energy, 151, p.107881_1 - 107881_13, 2021/02
被引用回数：14 パーセンタイル：89.48(Nuclear Science & Technology)日本原子力研究開発機構のMELT試験施設では、キログラムスケールの高温溶融ステンレス鋼とナトリウムとの相互作用を可視化するために、高分解能X線イメージングを用いている。本研究では、溶融ジェットの微粒化と急冷の定量的な評価のために、新しい画像処理ソフトウェアSPECTRAを開発した。X線による視野範囲を横切る溶融相の追跡と3次元再構成により、実験後に回収されたデブリの72%を検出することができた。融体と冷却材の界面に固化クラストが存在することが確認され、一方、熱的な微粒化に伴い急速な蒸気膨張が引き起こされた。本研究で観察された溶融ジェットの微粒化は、溶融ジェット内に取り込まれた冷却材が気化し、固化クラストの微粒化に十分な内部過圧を発生させたことで説明できる。この熱的な微粒化によって形成されたデブリは、粗いクラスト状の固化物とより細かい固化物から成る二つのピークを有する粒子径分布を示した。
仲吉 彬; Rempe, J. L.*; Barrachin, M.*; Bottomley, D.; Jacquemain, D.*; Journeau, C.*; Krasnov, V.; Lind, T.*; Lee, R.*; Marksberry, D.*; et al.
Nuclear Engineering and Design, 369, p.110857_1 - 110857_15, 2020/12
被引用回数：11 パーセンタイル：36.42(Nuclear Science & Technology)福島第一原子力発電所(1F)の各ユニットの燃料デブリの最終状態位置については、まだ多くは不明である。不確実性の低減に向けた最初のステップとして、OECD/NEAは、燃料デブリ分析予備的考察(PreADES)プロジェクトが立ち上げた。PreADESプロジェクトのタスク1の一環として、関連情報をレビューし、燃料デブリの状態の推定図の正確さを確認した。これは、将来の燃料デブリの分析を提案するための基礎となる。具体的にタスク1では2つのアクティビティを実施した。第一に、1Fでの廃止措置活動に資するTMI-2とチェルノブイリ原子力発電所4号機での重大事故の関連知識、プロトタイプ試験とホットセル試験の結果の知見を収集した。第二に、プラント情報とBSAFプロジェクトのシビアアクシデントコード分析からの関連知識が組み込まれている1F燃料デブリの原子炉内の状態に関する現状の推定図を見直した。この報告は、PreADESプロジェクトのタスク1の洞察に焦点を当て、1Fの将来の除染および廃止措置活動に情報を提供するだけでなく、シビアアクシデント研究、特にシビアアクシデントにより損傷した原子力サイトの長期管理に関する重要な視点を提供する。
Pellegrini, M.*; Herranz, L.*; Sonnenkalb, M.*; Lind, T.*; 丸山 結; Gauntt, R.*; Bixler, N.*; Morreale, A.*; Dolganov, K.*; Sevon, T.*; et al.
Nuclear Technology, 206(9), p.1449 - 1463, 2020/09
被引用回数：41 パーセンタイル：98.41(Nuclear Science & Technology)The OECD/NEA Benchmark Study at the Accident of Fukushima Daiichi Nuclear Power Station (BSAF) project, which started in 2012 and continued until 2018, was one of the earliest responses to the accident at Fukushima Daiichi. The project, divided into two phases addressed the investigation of the accident at Unit 1, 2 and 3 by Severe Accident (SA) codes until 500 h focusing on thermal-hydraulics, core relocation, Molten Corium Concrete Interaction (MCCI) and fission products release and transport. The objectives of BSAF were to make up plausible scenarios based primarily on SA forensic analysis, support the decommissioning and inform SA codes modeling. The analysis and comparison among the institutes have brought up vital insights regarding the accident progression identifying periods of core meltdown and relocation, Reactor Pressure Vessel (RPV) and Primary Containment Vessel (PCV) leakage/failure through the comparison of pressure, water level and CAMS signatures. The combination of code results and inspections (muon radiography, PCV inspection) has provided a picture of the current status of the debris distribution and plant status. All units present a large relocation of core materials and all of them present ex-vessel debris with Unit 1 and Unit 3 showing evidences of undergoing MCCI. Uncertainties have been identified in particular on the time and magnitude of events such as corium relocation in RPV and into cavity floor, RPV and PCV rupture events. Main uncertainties resulting from the project are the large and continuous MCCI progression predicted by basically all the SA codes and the leak pathways from RPV to PCV and PCV to reactor building and environment. The BSAF project represents a pioneering exercise which has set the basis and provided lessons learned not only for code improvement but also for the development of new related projects to investigate in detail further aspects of the Fukushima Daiichi accident.
仲吉 彬; Journeau, C.*; Rempe, J.*; Barrachin, M.*; Bottomley, D.; Nauchi, Y.*; Song, J. H.*
Proceedings of 2019 International Workshop on Post-Fukushima Challenges on Severe Accident Mitigation and Research Collaboration (SAMRC 2019) (USB Flash Drive), 6 Pages, 2019/11
In recognition of the broad international interest in learning from post-accident examinations and other activities related to the Fukushima Daiichi Nuclear Power Station (1F), Japan recommended, to the Organization for Economic Co-operation and Development/Nuclear energy Agency/Committee on Safety of Nuclear Installations (OECD/NEA/CSNI) in 2013, that they identified and followed up on opportunities to address safety research gaps. The CSNI set up the Senior Expert Group (SEG) on Safety Research Opportunities Post-Fukushima (SAREF). In 2016-2017, Preparatory Study on Analysis of Fuel Debris (PreADES) project was recommended by the SEG on SAREF as a near-term project. The PreADES project will summarize the collected knowledge and expertise of debris characterization and identify the needs for debris analyses that will most contribute to the decommissioning of 1F. The project also aims to improve the understanding of severe accidents and reactor safety assessments as well as creating appropriate and optimal methodologies for future debris sampling, retrieval, and storage. Consequently, the project provides important input for a future international project of sample examination based on long-term considerations. The PreADES project launched discussions among interested organizations at the preliminary meeting in July 2017 about the objectives, scope, output, and direction of the project. The contents of the PreADES project were agreed as the three following tasks: Task 1: Joint study on fuel debris' expected properties and characterization, Task 2: Identifying needs and major issues for future fuel debris sampling, retrieval, and analyses, Task 3: Planning of future international R&D framework. Currently, 4th meeting took place on July in Tokyo Japan. Task 1 is almost completed and Task 2 will be summarized soon.
仲吉 彬; 池内 宏知; 北垣 徹; 鷲谷 忠博; Bouyer, V.*; Journeau, C.*; Piluso, P.*; Excoffier, E.*; David, C.*; Testud, V.*
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05
A large-scale Molten Core-Concrete Interaction (MCCI) test (VF-U1) under the Fukushima Daiichi Nuclear Power Station (1F) conditions (core material composition, concrete, and decay heat) was conducted at the large MCCI test facility (VULCANO) owned by French Alternative Energies and Atomic Energy Commission (CEA) in France. About 50 kg of simulated debris was melted and brought into contact with concrete to erode concrete under 1F conditions. After cooling, the concrete test section (concrete and MCCI product) was dismantled. Main observations of the structure of solidified pool (crust, porosity, oxide/metal layer, etc.) and of the ablation are given. The technical results obtained herein are summarized, and they provide interesting knowledge that will help with the decommissioning of 1F.
Bouyer, V.*; Journeau, C.*; Haquet, J. F.*; Piluso, P.*; 仲吉 彬; 池内 宏知; 鷲谷 忠博; 北垣 徹
Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 13 Pages, 2019/03
Fuel debris removal is one of the most important processes for decommissioning a severely damaged nuclear power plant (NPP) such as Fukushima Daiichi NPP (1F). In order to develop relevant removal tools, characteristics of fuel debris are required. In the frame of a JAEA-CEA cooperation, a large-scale MCCI test was performed at the CEA/VULCANO facility using a prototypic metal and oxide corium representative from Fukushima Daiichi unit 1 conditions. Conclusions arising from the material analysis of the selected samples will be relevant for future dismantling operations. This paper deals with the experimental device and process, objective and initial conditions of this MCCI test, and ablation of the concrete quantified in term of volume, depths and velocities. The test section concrete, made with Japanese components, is siliceous with basaltic origin. The main objective of the test was to get a significant ablation leading to an ablation volume ratio of 1.6 in order to produce fuel debris with a composition corresponding to expected conditions in the damaged plant. On a phenomenological point of view, it must be noted that the concrete ablation was clearly anisotropic with a predominantly downwards ablation contrary to previous experiments with silica and limestone concrete.
Journeau, C.*; Bechta, S.*; Komlev, A.*; 倉田 正輝; 多木 寛; 松本 俊慶; Mohamad, A. B.; Barrachin, M.*; Quaini, A.*; Bottomley, D.*; et al.
no journal, ,
The OECD project TCOFF-2 (Thermodynamic Chemistry of Fission Products - Part 2) is an extension of the original project that was part of the near-term projects intended to support the Fukushima Dai-ichi decommissioning efforts. This second part continues to be mainly financed by Japanese Ministry (MEXT) with JAEA-CLADS support. TCOFF2 started in August 2022 and includes certain new material requirements compared to the first TCOFF project. These are increased emphasis on accident tolerant fuels (ATFs), certain actinides and fission products related to volatility/leachability but also certain combinations of the ceramic oxide systems important for MCCI behaviour. Following a PIRT review of phenomena in TCOFF 1, there was in TCOFF2 a Task 1 to prioritise current needs, particularly for relevance to severe accident phenomena and alternative materials (ATFs). This included a re-evaluation and ranking of the thermodynamic systems to make a Systems Identification and Ranking Table (SIRT) for the improvement of the thermodynamic database foreseen in TCOFF2. The further systems for evaluation were then proposed by the partners according to their particular requirements; this resulted in over 150 thermodynamic systems. In the mid-year meeting (June 2023) discussions were made to rationalise these into the most important 20 systems. The rationale for the reduction of systems to this limit will be explained in the talk, both those systems that were omitted as well as those finally included. These systems will then be used as a priority list of work for the experimental call to members that will be launched by the TCOFF-2 project towards the end of the year 2023 with the intention to initiate the first projects soon afterwards.
坪田 陽一; Laffolley, H.; Porcheron, E.*; Journeau, C.*; Delacroix, J.*; Guvar, C.*; Brackx, E.*; Lallot, Y.*; Bouland, A.*
no journal, ,
福島第一原子力発電所(1F)から燃料デブリを安全に取り出すためには、燃料デブリの切断により発生する放射性エアロゾルを定量的に評価する必要がある。そこで、In/Ex-Vesselの組成を模擬したウラン含有模擬燃料デブリを作製し、加熱により発生するエアロゾルの物理的・化学的特性を評価した。これらの結果に基づき、福島第一原子力発電所2号機の燃料デブリのレーザー切断(典型的な熱的切断工法)時に生成されるエアロゾルの同位体組成と放射能を推定した。プルトニウム、主にPu、Am、Cmが主要なアルファ核種であり、またPu、Cs-Ba、Sr-Yが主要なベータ核種であった。
Journeau, C.*; Bechta, S.*; Komlev, A.*; 倉田 正輝; 多木 寛; 松本 俊慶; Mohamad, A. B.; Barrachin, M.*; Quaini, A.*; Guneau, C.*; et al.
no journal, ,
Within the second phase of the OECD/NEA project on Thermodynamic Characterisation of Fuel Debris and Fission Products Based on Scenario Analysis of Severe Accident Progression (TCOFF-2), a Systems Identification and Ranking Table (SIRT), has been derived from the usual PIRT and adapted to chemical thermodynamics and material science. Firstly, a draft list of systems of interests has been distributed to TCOFF-2 partners. After review, a final list of 154 systems has been considered. Then system ranking has been carried out. Two series of figures of merit have been considered: importance for safety including the contribution to severe accident phenomena and to fission product behaviour and source term as well as needs for further R&D, based on the lack of existing data and the needs to improve relevant existing thermodynamic databases (mostly NUCLEA and TAF-ID). Each participating organization provided ranking of the systems for these Figure of Merits. Relevance to safety has been organized in 10 columns: Classical LWR / High BU, High enrichment fuel/ Fukushima Daiichi, TMI2, Chornobyl/Near-term ATF cladding (Cr-coated Zry)/ Near-term ATF cladding (FeCrAl) /Long-term ATF cladding (SiC) /Advanced fuels (UN, U-Si) /Advanced Modular Reactors / Post-Accident Leaching). Received rankings have then been agglomerated and averaged. A dedicated finalization meeting has been held in June 2023 in which the systems having the highest relevance and the largest needs for improvement have been considered. It must be noted that some systems for which there is a rather good knowledge, like (U,O, Zr) have not been selected, nor systems in which thermodynamic results are very hard to attain due to kinetics effect such as (U,H,O). After some system grouping, consensus have been reached on a list of 20 systems (13 linked to current or near-term fuels and 7 linked to longer-term fuels) having the highest priority.
松場 賢一; 江村 優軌; 神山 健司; Johnson, M.*; Journeau, C.*
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溶融燃料-冷却材相互作用(FCI)は、ナトリウム冷却高速炉(SFR)のシビアアクシデント時に発生する重要な現象である。FCIの発生に伴う溶融燃料の微粒化は、原子炉内下部構造物への溶融ジェットの衝突による溶融浸食を回避するために必要であるとともに、長期的な熱除去のためのデブリベッドの初期状態を決定づける要因である。SFR条件下で発生するFCIに関する理解を深め、評価モデルを開発するため、原子力機構(JAEA)とCEAは、JAEAが有するMELT試験施設で実施された溶融ステンレス鋼-ナトリウム相互作用試験で得られた結果の共同解釈を実施した。この共同研究を通じて、FCIの実験的解明とそれを支えるモデルにより、溶融ジェットの微粒化とデブリ形成のメカニズムを含むFCI挙動に関する理解を深めることができた。
Guvar, C.*; Faure, J.*; Testud, V.*; Roger, J.*; Domenger, R.*; Valette, R.*; Brackx, E.*; Bouyer, V.*; Journeau, C.*; Berlemont, R.*; et al.
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URASOLとダスト飛散率プロジェクトは、それぞれJAEAとCRIEPIがONET/CEA/IRSNと共同で主導し、熱または機械工具で切断された劣化ウラン酸化物を含む模擬燃料デブリからエアロゾルの生成と特性に関する基礎データを取得することを目的とする。ONET/CEA/IRSNが開発したプロセス全体では、模擬燃料デブリの仕込み組成決定と製作、切断用試料の準備、切断試験とオンラインでのエアロゾル測定、エアロゾルのサンプリング、模擬燃料デブリ製作時とエアロゾル生成試験後分析を実施可能である。本発表では、模擬燃料デブリ(Ex-Vessel組成のVF-U3)の製作時、及びそれらの加工時のエアロゾルの化学および微細構造分析に焦点を当てる。
