中村 秀夫; Bentaib, A.*; Herranz, L. E.*; Ruyer, P.*; Mascari, F.*; Jacquemain, D.*; Adorni, M.*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10
The WGAMA activity achievements have been published as technical reports, becoming reference materials to discuss innovative methods, materials and technologies in the fields of thermal-hydraulics, computational fluid dynamics (CFD) and severe accidents (SAs). The International Standard Problems (ISPs) and Benchmarks of computer codes have been supported by a huge amount of the databases for the code validation necessary for the reactor safety assessment with accuracy. The paper aims to review and summarize the recent WGAMA outcomes with focus on new advanced reactor applications including small modular reactors (SMRs). Particularly, discussed are applicability of major outcomes in the relevant subjects of passive system, modelling innovation in CFD, severe accident management (SAM) countermeasures, advanced measurement methods and instrumentation, and modelling robustness of safety analysis codes. Although large portions of the outcomes are considered applicable, design-specific subjects may need careful considerations when applied. The WGAMA efforts, experiences and achievements for the safety assessment of operating nuclear power plants including SA will be of great help for the continuous safety improvements required for the advanced reactors including SMRs.
Brumm, S.*; Gabrielli, F.*; Sanchez-Espinoza, V.*; Groudev, P.*; Ou, P.*; Zhang, W.*; Malkhasyan, A.*; Bocanegra, R.*; Herranz, L. E.*; Berda, M.*; et al.
Proceedings of 10th European Review Meeting on Severe Accident Research (ERMSAR 2022) (Internet), 13 Pages, 2022/05
The current HORIZON-2020 project on "Management and Uncertainties of Severe Accidents (MUSA)" aims at applying Uncertainty Quantification (UQ) in the modeling of Severe Accidents (SA), particularly in predicting the radiological source term of mitigated and unmitigated accident scenarios. Within its application part, the project is devoted to the uncertainty quantification of different severe accident codes when predicting the radiological source term of selected severe accident sequences of different nuclear power plant designs, e.g. PWR, VVER, and BWR. Key steps for this investigation are, (a) the selection of severe accident sequences for each reactor design, (b) the development of a reference input model for the specific design and SA-code, (c) the selection of a list of uncertain model parameters to be investigated, (d) the choice of an UQ-tool e.g. DAKOTA, SUSA, URANIE, etc., (e) the definition of the figures of merit for the UA-analysis, (f) the performance of the simulations with the SA-codes, and, (g) the statistical evaluation of the results using the capabilities, i.e. methods and tools offered by the UQ-tools. This paper describes the project status of the UQ of different SA codes for the selected SA sequences, and the technical challenges and lessons learnt from the preparatory and exploratory investigations performed.
Lebel, L. S.*; Morreale, A. C.*; Freitag, M.*; Gupta, S.*; Allelein, H.-J.*; Klauck, M.*; 孫 昊旻; Herranz, L. E.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Properly assessing pool scrubbing decontamination factors or radionuclide re-entrainment rates in a reactor safety analysis needs to be supported by a sufficiently robust experimental dataset, based on well qualified aerosol measurement techniques. A review of different pool scrubbing-related source term experiments has been conducted, along with a comparison of the measurement techniques that have been employed. In most areas, a fairly robust dataset exists to assess decontamination factors, but there is still a need to better understand some of the underlying aerosol mechanisms. The available dataset of re-entrainment experiments is smaller, and has gaps, for example, in pools with high velocity gas injections, or with re-flooded corium applications where the pool is undergoing film boiling. There are also many measurement techniques (e.g., cascade impactors, light scattering techniques, phase Doppler anemometry, etc.) that have different capabilities and are suitable for studying different aspects of the experiments. Linking the results that the techniques give, and how their results can ultimately be employed in safety analysis (including uncertainty quantification), is an important consideration in applying the results. This work was performed as a collaborative activity within the framework of the NUGENIA IPRESCA (Integration of Pool scrubbing Research to Enhance Source-term Calculations) project.
Marchetto, C.*; Ha, K. S*; Herranz, L. E.*; 廣瀬 意育; 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
After the Fukushima Daiichi accident of March 2011, one of the main concerns of the nuclear industry has been the research works for improving atmospheric radioactive release mitigation systems. Pool scrubbing is an important process in reactors that mitigates radioactive release. It is based on the injection of gases containing fission products through a water pool. Bubble hydrodynamics, as a result of gas injection and the associated water pool thermal-hydraulics, is an important aspect of the process since the bubble size, shape, velocity, etc. influence the fission product trapping at the bubble interface with the water. Computer codes dedicated to the pool scrubbing have been mainly developed in the 90's last century and modelling drawbacks have been identified in particular for bubble hydrodynamics. One of IPRESCA project objectives is to improve the pool scrubbing modelling. In order to highlight the main modelling issues, a benchmark exercise has been performed focusing on the bubble hydrodynamics. This benchmark, performed by nine organisations coming from six countries, aims at simulating a basic configuration, a single upward injector in ambient conditions, experimentally characterized in the RSE tests carried out in the European PASSAM project. In this paper, a short description of the code modelling and a comparison between the code results and the experimental data are presented and discussed. Then, outcomes from the benchmark result analysis and proposals of improvements are emphasized.
Lind, T.*; Herranz, L. E.*; Sonnenkalb, M.*; 丸山 結
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 15 Pages, 2022/03
The accident progression and fission product release from the three damaged units of the Fukushima Daiichi Nuclear Power Plant were systematically investigated in the OECD/NEA BSAF project phases 1 and 2. As a result of those investigations, a good progress was achieved in establishing defendable accident scenarios and the corresponding fission product releases to the environment. Nonetheless, there are some areas requiring further work, particularly concerning fission product behavior. They are addressed in the OECD/NEA project "Analysis of Information from Reactor Buildings and Containment Vessels of Fukushima Daiichi NPS" (ARC-F). Based on the outcome of the BSAF project, several focus areas were selected for further investigations in the ARC-F project, one of them being the behavior of fission products and source term. In this paper, five topics which were ranked with a high significance as open issues based on the BSAF project regarding fission product behavior are discussed: i) fission product speciation, ii) iodine chemistry, iii) pool scrubbing, iv) fission product transport and behavior in the buildings, and v) uncertainty analysis and variant calculations. Significant progress has been made in these five topics in the ARC-F project. In this paper, background is given for choosing these topics for specific investigations based on the outcome of the BSAF project. The topics are described and the approach to study them in the ARC-F given along with some exemplary, preliminary results. Finally, the readers' attention is drawn to open issues which are not included in the ARC-F work scope and could need further attention.
Gupta, S.*; Herranz, L. E.*; Lebel, L. S.*; Sonnenkalb, M.*; Pellegrini, M.*; Marchetto, C.*; 丸山 結; Dehbi, A.*; Suckow, D.*; Krkel, T.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Pool scrubbing is a major topic in water cooled nuclear reactor technology as it is one of the means for mitigating the source-term to the environment during a severe accident. Pool scrubbing phenomena include coupled interactions between bubble hydrodynamics, aerosols and gaseous radionuclides retention mechanisms under a broad range of thermal-hydraulic conditions as per accident scenarios. Modeling pool scrubbing in some relevant accident scenarios has shown to be affected by substantial uncertainties. In this context, IPRESCA (Integration of Pool scrubbing Research to Enhance Source-term CAlculations) project aims to promote a better integration of international research activities related to pool scrubbing by providing support in experimental research to broaden the current knowledge and database, and by supporting analytical research to facilitate systematic validation and model enhancement of the existing pool scrubbing codes. The project consortium includes more than 30 organisations from 15 countries involving research institutes, universities, TSOs, and industry. For IPRESCA activities, partners join the project with in-kind contributions. IPRESCA operates under NUGENIA Technical Area 2/SARNET (Severe Accident) - Sub Technical Area 2.4 (Source-term). The present paper provides an introduction and overview of the IPRESCA project, including its objectives, organizational structure and the main outcomes of completed activities. Furthermore, key activities currently ongoing or planned in the project framework are also discussed.
Arokiaswamy, J. A.*; Batra, C.*; Chang, J. E.*; Garcia, M.*; Herranz, L. E.*; Klimonov, I. A.*; Kriventsev, V.*; Li, S.*; Liegeard, C.*; Mahanes, J.*; et al.
IAEA-TECDOC-2006, 380 Pages, 2022/00
Lind, T.*; Pellegrini, M.*; Herranz, L. E.*; Sonnenkalb, M.*; 西 義久*; 玉置 等史; Cousin, F.*; Fernandez Moguel, L.*; Andrews, N.*; Sevon, T.*
Nuclear Engineering and Design, 376, p.111138_1 - 111138_12, 2021/05
Khatib-Rahbar, M.*; Barrachin, M.*; Denning, R.*; Gabor, J.*; Gauntt, R.*; Herranz, L. E.*; Hobbins, R.*; Jacquemain, D.*; 丸山 結; Metcalf, J.*; et al.
NUREG/CR-7282, ERI/NRC 21-204 (Internet), 160 Pages, 2021/04
The U.S. Nuclear Regulatory Commission (NRC) is preparing to accept anticipated licensing applications for the commercial use of accident tolerant fuel (ATF) in commercial nuclear power plants in the United States. It is the objective of the NRC to evaluate the effects of ATF designs on severe accident behavior, and to determine potential changes to the NRC severe accident analysis computer codes that would simulate plant conditions using ATFs commensurate with the accuracy in accident analyses involving conventional fuels. This report documents the development of Phenomena Identification and Ranking Tables (PIRTs) for near-term ATFs under severe accident conditions in light water reactors (LWRs). The PIRTs were developed by a panel of experts for various near-term ATF design concepts (i.e., FeCrAl cladding, zirconium alloy cladding coated with chromium, and CrO dopants in uranium dioxide fuels) in addition to the impacts from fuel enrichment and burnup. Panel members also considered the severe accident implications of the longer-term ATF concepts. The main figures-of-merit considered in this ranking process are the amount of fission products released into the containment and the quantity of combustible gases generated during an accident. Special focus is given to whether existing severe accident codes and models would be sufficient as applied to LWRs employing these fuels, and whether additional experimental studies or model development would be warranted.
Herranz, L. E.*; Pellegrini, M.*; Lind, T.*; Sonnenkalb, M.*; Godin-Jacqmin, L.*; Lpez, C.*; Dolganov, K.*; Cousin, F.*; 玉置 等史; Kim, T. W.*; et al.
Nuclear Engineering and Design, 369, p.110849_1 - 110849_7, 2020/12
Sonnenkalb, M.*; Pellegrini, M.*; Herranz, L. E.*; Lind, T.*; Morreale, A. C.*; 神田 憲一*; 玉置 等史; Kim, S. I.*; Cousin, F.*; Fernandez Moguel, L.*; et al.
Nuclear Engineering and Design, 369, p.110840_1 - 110840_10, 2020/12
Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; 中村 秀夫
Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09
WGAMA started on Dec. 31st 1999 to assess and strengthen the technical basis needed for the prevention, mitigation and management of potential accidents in NPP and to facilitate international convergence on safety issues and AM analyses and strategies. WGAMA addresses reactor thermal-hydraulics (Thys), in-vessel behavior of degraded cores, containment behavior and protection, and FP release, transport, deposition and retention, for both current and advanced reactors. This paper summarizes such WGAMA contributions in Thys, CFD and severe accidents, which include the Fukushima-Daiichi accident impacts on the WGAMA activities and their substantial outcomes. Around 50 technical reports have become reference in the related fields, which appear in References. Recommendations in these reports include further research, some of which have given rise to the joint projects conducted or underway within the OECD framework. Ongoing WGAMA activities are numerous and a number of them are to be launched in the near future, which are shortly mentioned too.
Allelein, H.-J.*; Auvinen, A.*; Ball, J.*; Gntay, S.*; Herranz, L. E.*; 日高 昭秀; Jones, A. V.*; Kissane, M.*; Powers, D.*; Weber, G.*
NEA/CSNI/R(2009)5, 388 Pages, 2009/12
The TMI accident in 1979 motivated an interest in LWR source terms and resulted in the production of a supplement to the first state of the art report (SOAR) which concentrated on LWR aerosol issues. The second SOAR dealt with primary system FP release and transport that covers vapor the condensation on aerosols and aerosol agglomeration. The present third SOAR was prepared focusing on aerosol behavior in both the primary circuit and in containment such as mechanical resuspension, impact of chemistry, re-vaporization of deposits, charge effect, removal by spray, hydrogen-burn effects on suspended aerosols, penetration of aerosols through leak paths and so on. A large number of probabilistic safety analysis (PSA level 2) plant studies have been performed around the world, frequently involving aspects of aerosol behavior. This report provides some examples, including sensitivity studies that demonstrate the impact of aerosol-related processes.