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Tsubota, Yoichi; Porcheron, E.*; Journeau, C.*; Delacroix, J.*; Suteau, C.*; Lallot, Y.*; Bouland, A.*; Roulet, D.*; Mitsugi, Takeshi
Proceedings of International Conference on Environmental Remediation and Radioactive Waste Management (ICEM2023) (Internet), 6 Pages, 2023/10
In order to safely remove fuel debris from the Fukushima Daiichi Nuclear Power Station (1F), it is necessary to quantitatively evaluate radioactive airborne particulate generated by the cutting of nuclear fuel debris. We fabricated Uranium-bearing simulated fuel debris (SFD) with In/Ex-Vessel compositions and evaluated the physical and chemical properties of aerosols generated by heating the SFDs. Based on these results, we estimated the isotopic composition and radioactivity of aerosols produced when 1F-Unit2 fuel debris is laser cut, which is a typical example of a heating method. Plutonium, mainly Pu,Am, and Cm were found to be the alpha nuclide, and Pu, Cs-Ba, and Sr-Y were found to be the beta nuclide of interest.
Porcheron, E.*; Journeau, C.*; Delacroix, J.*; Berlemont, R.*; Bouland, A.*; Lallot, Y.*; Tsubota, Yoichi; Ikeda, Atsushi; Mitsugi, Takeshi
Proceedings of International Conference on Environmental Remediation and Radioactive Waste Management (ICEM2023) (Internet), 5 Pages, 2023/10
Results of the URASOL project aimed at evaluating the generation and dispersion of radioactive aerosols during the cutting of fuel debris, a key issue in the decommissioning of the damaged reactors at the Fukushima Daiichi Nuclear Power Station (1F), are presented in this report. Characterization of aerosols generated during heating and mechanical cutting of simulated fuel debris in terms of mass concentration, real-time number density, mass-based particle size distribution, morphology, and chemical properties is reported. In the heating tests, an increase in particle size with increasing temperature was observed, and in terms of particle number density, the case using depleted uranium simulated fuel debris had a smaller number density than the test using Hf-containing simulated fuel debris. In mechanical cleavage, the aerodynamic median mass diameter of the aerosol was almost the same for the radioactive and non-radioactive samples (about 3.74.4 m).
Sun, Haomin; Leblois, Y.*; Gelain, T.*; Porcheron, E.*
Journal of Nuclear Science and Technology, 59(11), p.1356 - 1369, 2022/11
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)In severe accident scenarios of PWR, containment spray can be employed to washout the aerosol of radioactive materials, retaining them in the containment. Therefore, it is crucial to correctly predict the washout efficiency for safety assessment. For a PWR, a high spray coverage ratio ( 84%-95%) is required. However, experimental studies on the washout with such a high coverage ratio in a large vessel are quite limited. To understand such a washout phenomenon for model development, aerosol washout experiments are performed in a large vessel with not only aerosol measurements but also spray droplet characterizations. The spray coverage ratios are experimentally confirmed to be compatible with a real PWR. The washout features are investigated in detail. The model in MELCOR is examined using the measured aerosol removal rate, showing the removal rate tendency against particle diameters being reproduced. Although a significant underestimation occurs for large particles, a satisfactory agreement is obtained for smaller ones (0.52 m in diameter) corresponding to the minimum removal rate and around.
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
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 project, which is undertaken by a French consortium consisting of ONET Technologies, CEA, and IRSN for JAEA/CLADS, is dedicated to acquiring basic scientific data on the generation and characteristics of radioactive aerosols from the thermal or mechanical processing of fuel debris simulant. Heating process undertaken in the VITAE facility simulates some representative conditions of thermal cutting by LASER. For mechanical cutting, the core boring technique is implemented in the FUJISAN facility. Fuel debris simulants have been developed for inactive and active trials. The aerosols are characterized in terms of mass concentration, real time number concentration, mass size distribution, morphology, and chemical properties. The chemical characterization aims at identifying potential radioactive particles released and the associated size distribution, both of which are important information for assessing possible safety and radioprotection measures during the fuel debris retrieval operations at 1F.
Marchetto, C.*; Ha, K. S*; Herranz, L. E.*; Hirose, Yoshiyasu; Jankowski, T.*; Lee, Y.*; Nowack, H.*; Pellegrini, M.*; Sun, X.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 17 Pages, 2022/03
Gunji, Satoshi; Tonoike, Kotaro; Clavel, J.-B.*; Duhamel, I.*
Journal of Nuclear Science and Technology, 58(1), p.51 - 61, 2021/01
Times Cited Count:1 Percentile:12.16(Nuclear Science & Technology)The new critical assembly STACY will be able to contribute to the validation of criticality calculations related to the fuel debris. The experimental core designs are in progress in the frame of JAEA/IRSN collaboration. This paper presents the method applied to optimize the design of the new STACY core to measure the criticality characteristics of pseudo fuel debris that simulated Molten Core Concrete Interaction (MCCI) of the fuel debris. To ensure that a core configuration is relevant for code validation, it is important to evaluate the reactivity worth of the main isotopes of interest and their k sensitivity to their cross sections. In the case of the fuel debris described in this study, especially for the concrete composition, silicon is the nucleus with the highest k sensitivity to the cross section. For this purpose, some parameters of the core configuration, as for example the lattice pitches or the core dimensions, were adjusted using optimization algorithm to find efficiently the optimal core configurations to obtain high sensitivity of silicon capture cross section. Based on these results, realistic series of experiments for fuel debris in the new STACY could be defined to obtain an interesting feedback for the MCCI. This methodology is useful to design other experimental conditions of the new STACY.
Sun, Haomin; Porcheron, E.*; Magne, S.*; Leroy, M.*; Dhote, J.*; Ruffien Ciszak, A.*; Bentaib, A.*
Proceedings of OECD/NEA Specialist Workshop on Advanced Measurement Method and Instrumentation for enhancing Severe Accident Management in an NPP addressing Emergency, Stabilization and Long-term Recovery Phases (SAMMI 2020) (Internet), 10 Pages, 2020/12
Bentaib, A.*; Janin, T.*; Porcheron, E.*; Magne, S.*; Leroy, M.*; Dhote, J.*; Ruffien Ciszak, A.*; Sun, Haomin
Proceedings of OECD/NEA Specialist Workshop on Advanced Measurement Method and Instrumentation for enhancing Severe Accident Management in an NPP addressing Emergency, Stabilization and Long-term Recovery Phases (SAMMI 2020) (Internet), 6 Pages, 2020/12
Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; Nakamura, Hideo
Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09
Times Cited Count:3 Percentile:11.26(Nuclear Science & Technology)Sun, Haomin; Leblois, Y.*; Gelain, T.*; Porcheron, E.*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 11 Pages, 2020/08
Chadwick, M. B.*; Capote, R.*; Trkov, A.*; Herman, M. W.*; Brown, D. A.*; Hale, G. M.*; Kahler, A. C.*; Talou, P.*; Plompen, A. J.*; Schillebeeckx, P.*; et al.
Nuclear Data Sheets, 148, p.189 - 213, 2018/02
Times Cited Count:66 Percentile:98.06(Physics, Nuclear)The CIELO collaboration has studied neutron cross sections on nuclides that significantly impact criticality in nuclear facilities - U, U, Pu, Fe, O and H - with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform well in integral simulations of criticality. This report summarizes our results and outlines plans for the next phase of this collaboration.
Izawa, Kazuhiko; Tonoike, Kotaro; Leclaire, N.*; Duhamel, I.*
Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.965 - 976, 2015/09
In the post-Fukushima prospect, Japan Atomic Energy Agency, JAEA, intends to re-open its Static Experiment Critical Facility STACY, in order to contribute to the validation of fuel debris calculations in conditions close to those affected by the Fukushima Daiichi Nuclear Power Station accident. The reactor core of STACY is being modified by JAEA from homogeneous solution fuel core to heterogeneous pin-type fuel core. The first step of critical experiments is to establish reference experiments involving lattices of UO rods without fuel debris. A design of such experiments is in progress for the widest range of moderation ratios in the frame of JAEA/IRSN collaboration. For that purpose, perturbation calculation technique and stochastic geometry modeling technique were used with the MCNP5 and the JENDL-4.0 evaluated nuclear data library. The proposed paper will discuss all the aforementioned issues.
Takano, Kazuya; Mori, Tetsuya; Kishimoto, Yasufumi; Hazama, Taira
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 13 Pages, 2014/09
This paper describes details of the IAEA/CRP benchmark calculation by JAEA on the control rod withdrawal test in the Phenix End-of-Life Experiments. The power distribution deviation by the control rod insertion/withdrawal, which is the major target of the benchmark, is well simulated by calculation. In addition to the CRP activities, neutron and photon transport effect is evaluated in the nuclear heating calculation of the benchmark analysis. It is confirmed that the neutron and photon transport effect contributes to the improvement of the absolute power calculation results in the breeder blanket region.
Pascal, V.*; Prulhire, G.*; Vanier, M.*; Fontaine, B.*; Devan, K.*; Chellapandi, P.*; Kriventsev, V.*; Monti, S.*; Mikityuk, K.*; Chenu, A.*; et al.
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 16 Pages, 2014/09
no abstracts in English
Monti, S.*; Toti, A.*; Stanculescu, A.*; Pascal, V.*; Fontaine, B.*; Herrenschmidt, A.*; Prulhiere, G.*; Vanier, M.*; Varaine, F.*; Vasile, A.*; et al.
IAEA-TECDOC-1742, 247 Pages, 2014/06
Pascal, V.*; Prulhire, G.*; Fontaine, B.*; Devan, K.*; Chellapandi, P.*; Kriventsev, V.*; Monti, S.*; Mikityuk, K.*; Semenov, M.*; Taiwo, T.*; et al.
Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 11 Pages, 2013/04
The control rod withdrawal test was one of the various Phenix End-of-Life tests performed in 2009. The main goal was to determine the impact of a rod insertion and/or extraction on the radial power distribution in the fissile core at nominal power. The framework of the Technical Working Group on Fast Reactors (TWG-FR) activities in IAEA, decided to launch a Coordinated Research Project (CRP), devoted to benchmarking analyses on the test. The CRP was performed by experts coming from CEA, ANL, IGCAR, IPPE, IRSN, JAEA, KIT and PSI. After a short description of the test conducted in the Phenix reactor, this paper presents some results obtained in the course of the CRP with special emphasis on control rod efficiencies and power deformation by subassemblies. The paper also discusses the discrepancies found when comparing calculated results with experimental data as well as some preliminary conclusions on the source of these discrepancies.
Trgours, N.*; Sugiyama, Tomoyuki
DPAM-SEMCA-2010-288, 73 Pages, 2010/11
The JAEA and IRSN are planning an experimental program to clarify the dynamics in fission gas release from fuel pellets under reactivity-initiated accident (RIA) conditions. In the Fission Gas Dynamics (FGD) test, high burnup fuel pellets in a rigid chamber is pulse-irradiated in the NSRR and the history of fission gas release is evaluated by transient measurements of the pressure and temperature in the chamber. In order to validate the chamber design and test conditions, preparatory calculations were performed on the high burnup UO and MOX fuels, which had been tested in the NSRR, using the transient fuel behavior analysis code SCANAIR. The calculations provided informative results for the chamber design; the newly developed pressure sensor, which uses a linear variable differential transformer (LVDT), has an enough quick response, the thermal insulation around the chamber can be omitted because of the large heat capacity of the chamber wall, and so on.
Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken
Journal of Nuclear Science and Technology, 47(4), p.396 - 410, 2010/04
Times Cited Count:7 Percentile:45.04(Nuclear Science & Technology)Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, J.*
Journal of Nuclear Science and Technology, 46(11), p.1049 - 1058, 2009/11
Times Cited Count:17 Percentile:72.88(Nuclear Science & Technology)Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, J.*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 13 Pages, 2009/10
Georgenthum, V.*; Sugiyama, Tomoyuki
DPAM-SEMCA-2008-330, 36 Pages, 2009/05
The JAEA and IRSN are planning an experimental program to clarify the dynamics in fission gas release from fuel pellets under reactivity-initiated accident (RIA) conditions. In the Fission Gas Dynamics (FGD) test, high burnup fuel pellets are pulse-irradiated in the NSRR, and dynamic behavior of fission gas release and its contribution to the mechanical load on the cladding are evaluated. This report presents the objectives of the FGD programme, progress of the test preparation, pending questions and preliminary calculation results of high burnup fuel tests.