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芝 知宙; 冠城 雅晃; 能見 貴佳; 鈴木 梨沙; 小菅 義広*; 名内 泰志*; 高田 映*; 長谷 竹晃; 奥村 啓介
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 3 Pages, 2022/10
A technique that can easily determine the presence of nuclear material in removed object from Fukushima Daiichi Nuclear Power Plant site is important from the viewpoint of sorting fuel debris from radioactive waste. In the case of fresh uranium, the amount of nuclear material in the waste generated from nuclear facilities can be determined by measuring 1001 keV gamma-rays emitted by 
Pa, which is a daughter nuclide of 
U. However, it has been pointed out that such gamma-ray measurement cannot be used for fuel debris that contains a large portion of fission products (FPs) emitting various energies of gamma-rays. In this study, we focus on prompt fission gamma-rays that are directly emitted from nuclear materials and those energy exists in a higher energy region than those of FPs, and aim to measure them in simple manners.
Pshenichnikov, A.; 倉田 正輝; 永江 勇二
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 4 Pages, 2022/10
CLADS-MADE-04は、下部コア領域での溶融伝播挙動の理解を目的としたシリーズの次のテストである。この寄稿では、電子プローブ マイクロアナライザー(EPMA)によって調査された金属破片の微細構造を含む試験後分析の最近の結果について説明する。テスト中、制御棒ブレードの溶融は、比較的ゆっくりと(数cm/分)急激な強い熱放出の波で発生し、最も高温の領域から、劣化しつつある制御棒ブレードとチャネルボックスに沿って下方に広がり、ジルカロイ-4で作られた壁を消費した。サンプル支持板にも大きな損傷が発生した。このような金属リッチ破片の微細構造の調査により、強化された局所コア劣化のメカニズムを理解できるようになる。EPMAによる相同定を徹底した上で、放熱性が高く周囲への拡散の可能性があることを確認する必要がある。Fe-B共晶デブリとZr-Fe共晶デブリの違いについて概説する。これは、下部炉心プレートのメルトスルーと、下部プレナムへのZr-Fe溶融材料の進行の可能性を理解するために特に重要である。
鈴木 健太; 八代 大*; 川端 邦明
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 4 Pages, 2022/10
This paper describes a development of PHITSPlugin for the radiation behavior calculation. The developed plugin calculates a dose distribution in conjunction with Choreonoid which is a physical simulator. It was developed to contribute to estimate an integral radiation dose on the robots. We discuss a procedure for calculating the dose distribution. Also, we demonstrate to calculate the dose distribution by utilizing experimental examples.
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。
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。
坂東 大都*; 佐々木 凌太郎*; 福田 貴斉*; 山路 哲史*; 山下 拓哉
no journal, ,
As one of the three tasks of "Project of Decommissioning, Contaminated Water and Treated Water Management (Development of Analysis and Estimation Technologies for Characterization of Fuel Debris) (Development of Estimation Technologies of RPV Damaged Condition, etc.)", this study presents evaluation of the fuel debris behavior below the damaged RPV lower head boundary of Fukushima Daiichi Nuclear Power Station (1F) Unit-2. The focus of the study is to evaluate the debris behavior at the time of / after the failure of the RPV boundary. It is expected to provide more comprehensive understanding of the precedingly obtained muon image, which seemed to indicate that a large amount of highly-dense materials distributed between the RPV lower head and the thermal insulation structures just below the RPV. The Moving Particle Semi-implicit (MPS) method is being developed to evaluate the fuel debris behavior in/under the actual plant geometry and conditions. The melt behavior analysis code, based on the MPS method, is being developed to analyze the following two debris behaviors. Firstly, the debris discharge behavior from penetration tube structures is analyzed. The solidified debris blocks are represented by rigid bodies, using the Passively Moving Solid (PMS) model with consideration of decay heat of the oxidic fuel debris. The relocations of the oxidic debris involving melting of the surrounding metallic debris and the penetration tube wall structure are analyzed. Secondly, the melt behavior on / through the multi-layered thermal insulation structures below the RPV is analyzed. The discharged melt from the RPV boundary may freeze on the insulation plate, depending on the thermal condition in the pedestal and the discharged melt history.
柳澤 華代; 松枝 誠; 古川 真*; 平田 岳史*; 高貝 慶隆*
no journal, ,
LA-ICP-MS法はXRF, EPMA, SIMSといった従来法よりも高感度に元素分析と同位体マッピングが可能である。しかし、Sr-90の場合はスペクトルの干渉や隣接する同位体のテーリングなどにより困難であった。特に、燃料被覆管や土壌中に存在するZr-90による同位体干渉が問題となっていた。本研究では、Zr-90を含む干渉の寄与を低減するために、ICP-タンデム質量分析計とダイナミックリアクションセルの組合せを採用した。本法により、分析領域(100
100
mm四方)から得られた検出限界は0.64mBqであった。m/z90uの測定バックグラウンド強度(1.6
2.2cps)は、種々の妨害元素(Ni: 
2,084mg/kg, Ge: 21mg/kg, Y: 23,004mg/kg, Zr: 175mg/kg)が存在しても大きく変化しなかった。また、Sr-88からSr-90への寄与は2.6e-9であった。本法はSr-90を含む微量元素の高感度マッピング分析を実現するものである。
佐藤 拓未; 山下 拓哉; 下村 健太; 永江 勇二
no journal, ,
The reaction between the molten metallic debris pool and the structural materials of reactor pressure vessels (RPV) is important in understanding the RPV failure behavior. In this study, the ELSA (Experiment on Late In-vessel Severe Accident Phenomena)-1 test, which focuses on the damage caused by the eutectic melting of the liquid metal pool and control rod drive (CRD) structures, was conducted. A test sample simulating the CRD structure at the lower head was fabricated and loaded with Fe-Zr alloy as the simulated metal debris. The sample was gradually heated to about 1400 
C using the LIESAN test facility, and in-situ observation was performed using a video camera. The test results showed that the CRD structural material reacted with the metal debris and melted at about 1050-1250 C, which was lower than the melting point of the CRD itself. It was also observed that the molten material flowed into the CRD, suggesting that the CRD structure was preferentially damaged during the severe accident.
山田 大地; 阿部 浩之*; 川端 邦明
no journal, ,
In emergency response and decommissioning at Fukushima Daiichi Nuclear Power Station (FDNPS), remote controlled robot is required to work in harmful places instead of human workers. However, remote robot operation is not easy. In order to achieve a mission by remotely robot operation, the robot operator should understand what and how much he/she is capable to do by remote operation with the robot. Thus, a method for evaluation of robot capability is needed. Many of the missions by remote robot operation in FDNPS accomplish main objectives, while, few problems occurred in some missions. The experience from such problems in FDNPS have significant value for practical remote robot operation. The aim of this research is to develop a robot testing method based on cases of FDNPS remote operations with arm mounted robots. We survey FDNPS remote operations with arm mounted robots, and decide to develop a testing method to evaluate the robot capability for approaching a target object over an obstacle by a robot arm. We develop a test field for evaluation the capability and the test procedure to measure the metric to describe the capability using the test field. Moreover, we conduct actual tests and discussions with a team engaged in nuclear emergency response robot maintenance and training to make it more practical.
高畠 容子; 駒 義和
no journal, ,
福島第一原子力発電所事故により、放射性核種が拡散した。廃炉作業により多量の放射性固体廃棄物が発生している。固体廃棄物の放射能インベントリを求める必要がある。コバルト60はキー核種として利用されることがあり、福島第一原子力発電所で発生する固体廃棄物にも適用する可能性を検討する必要がある。本研究では、コバルト60と放射性核種の関係を議論した。物理化学的性質や核種移行過程に関わらず、コバルト60はいくつかの核種と高い相関があった。高い相関性からコバルト60はスケーリングファクター法のキー核種として利用できる可能性がある。
佐藤 優樹
no journal, ,
In this presentation, the concept of an integrated Radiation Imaging System (iRIS) and its demonstration will be presented. The iRIS concept combines radiation measurement technology with elemental technologies such as SLAM, data visualization, robotics, and xR to remotely measure the distribution of radioactive substances and dose rate information, and visualize it in 3D. In a radiation environment such as the Fukushima Daiichi Nuclear Power Station (FDNPS), visualization of these radiation information will contribute to the reduction of worker exposure and work planning. As one of the demonstration examples, a visualization test of highly radioactive contamination was conducted near the Unit 1/2 exhaust stack of the FDNPS. Using a combination of a compact Compton camera and SLAM technology based on 3D-LiDAR, the author succeeded in drawing a 3-D map of the entire working environment visualizing the high level of radioactive contamination in the lower part of the exhaust stack using data acquired while moving. In addition, information on the dose rate along the system's moving trajectory was recorded on the 3-D map. In addition, by importing the 3-D map into a VR system and using a commercially available VR head-mounted display, users can experience a working environment that displays a highly contaminated part regardless of their location. Furthermore, a technology to visualize images of radioactive substances acquired by gamma-ray imagers in real space using AR technology has been developed, and verification tests are underway. Currently, the author is constructing a system that enables verification of the effectiveness of shielding and decontamination against radioactive substances in a virtual space by using the 3-D map that visualizes information on highly contaminated part and dose rates. In other words, this is an attempt to build a digital twin of the radiation work environment based on the iRIS concept.
舟木 健太郎
no journal, ,
This lecture overviews ongoing R&D efforts of the Japan Atomic Energy Agency for the decommissioning of Tokyo Electric Power Company Holdings, Inc.'s Fukushima Daiichi Nuclear Power Station, particularly focused on R&D outcomes related to fuel debris retrieval and waste management.
particle diameters by an alpha particle imaging detector森下 祐樹; 佐川 直貴; 高田 千恵; 百瀬 琢麿; 高崎 浩司
no journal, ,
Alpha emitting radionuclides such as plutonium (Pu) isotopes require special consideration in terms of internal exposure. The evaluation of the diameters (activity median aerodynamic diameter (AMAD)) of PuO particles is very important for internal exposure dose evaluation. In this study, we propose a new method to determine the PuO particle diameter from the energy spectrum obtained by the developed alpha particle imaging detector. PuO particles with different diameters were modeled by Monte Carlo simulation. The change in the shape of the energy spectrum for each particle diameter was evaluated. Two different patterns were modeled: (1) the case of 
PuO, and (2) the case of PuO (including isotopic composition of Pu). The value of the energy resolution of the simulation was based on the measured value of the detector energy resolution. Multiple regression analysis was performed to determine the PuO particle diameter from the obtained parameters. The simulation results showed that the radioactivity (measured counts) was lower when the PuO particle size was smaller. In addition, the energy spectrum was sharper due to the smaller effect of self-absorption of the PuO particles themselves. On the other hand, when the PuO diameter was large, the radioactivity (measured counts) was high. In addition, the energy spectrum became broad and the position of the peak shifted to the lower side. Using the obtained parameters (measured radioactivity, peak counts, peak position, and peak width), multiple regression analysis was performed to evaluate the PuO diameter. The proposed method was compared with the conventional method and found to be in agreement.
Cantarel, V.; 山岸 功
no journal, ,
Cementitious structure repair is one of the potential applications of geopolymer material. The context of decommissioning and decontamination efforts gives rises to several related applications. For instance, geopolymer could be used to manage cement waste, create seals to isolate a section of the decommissioning area, or cover contaminated concrete to protect workers. In that context, the property of geopolymer for immobilization of nuclides such as cesium grants additional benefits. Moreover, in maritime environments, technical studies showed low permeability properties of geopolymer coated concretes. However, the origin of that property was not fully explained and only loosely attributed to the interface between the two materials. The purpose of this study was to investigate the interface between ordinary Portland cement and geopolymer. Other materials, such as activated slags, use the same type of alkaline solution as geopolymer. We wanted to know if the properties achieved at the interface with concrete are related to the aggressive alkaline solution or specific to geopolymer. To this end, cement was either dipped in alkaline silicate solution for various duration or embedded in geopolymer. The immersion in activating solution induced a localized carbonation at the surface of the cement similar to what is observed with waterglass treatment of ordinary Portland cements. Contrary to standard carbonation, the carbonation progresses in a root-like pattern inward, following the CSH gel into the cement. When embedded in geopolymer, a 0.03 mm thick transition zone is formed at the interface with cement. The elemental evolution throughout the transition zone indicates that surface porosity of the cement is plugged by nanoaggregates of alumino-silicate. This phenomenon can explain both that the carbonation does not occur during or after the geopolymerization and the low permeability observed for geopolymer coated cements in other studies.