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Kawaguchi, Munemichi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
In decommissioning sodium-cooled fast reactors, the operators can be exposed to radiation during dismantling of the cold trap equipment (C/T). The C/T has higher dose equipment because the C/T trapped the tritium of the fission product during the operation to purify the sodium coolant. In this research, the thermal decomposition temperature and rate of sodium hydride were measured as the fundamental research for improvement of the thermorlysis method prior to the dismantling. We measured the thermal decomposition temperature and rate using sodium hydride powder (95.3%, Sigma-Aldrich) in Al
O
crucible with TG-DTA (STA2500, NETASCH Japan). The heating rates were set to 
= 2.0, 5.0, 10.0, 20.0 K/min, and the weight decrease was measured. The thermal decomposition temperature and rate were obtained from the temperature of the onset of the weigh decrease and the Kissinger plot, respectively. Furthermore, we set to the thermal decomposition temperature of around 600 K, and the weight decreasing rates were measured. The change of the chemical composition of the sodium hydride with heating (from NaH to Na) was measured with X-Ray Diffraction (XRD) analysis. As a result, the thermal decomposition occurred at around 600 K, and the almost all hydrogen in the sodium hydride released within 1 h. The thermal decomposition rate strongly depended on the heating temperature.
Shimada, Asako; Sawaguchi, Takuma; Takeda, Seiji
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 5 Pages, 2020/08
no abstracts in English
Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
Kang, Z.; Okuda, Yukihiko; Nishida, Akemi; Tsubota, Haruji; Li, Y.
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
Plenty of researches have been carried out to establish a rational assessment method for nuclear power plants against local damage caused by accidental projectile impact. Most of the empirical formulas have been proposed for quantitatively investigating the local damage to reinforced concrete (RC) structures caused by rigid projectile impact. These formulas have been derived on the basis of impact tests performed perpendicular to the target structure, while few impact tests oblique to the target structures have been studied. The final objective of this study is to propose a new formula for evaluating the local damage to RC structures caused by oblique impact based on experimental and simulation results. At present, we have validated an analytical method via comparison with experimental results and have conducted simulation analyses of oblique impact assessments on RC slab using various projectiles with flat nose shape by this method. In this study, the same analytical method will be applied to investigate the perforation damage to RC slab subjected to oblique impact by projectiles with hemispherical nose shape. In this paper, the effects of projectile's nose shape on the local damage of RC slab, the residual velocity of projectile and the time history of energy transmission will be discussed.
Wang, Z.; Iwasawa, Yuzuru; Sugiyama, Tomoyuki
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 12 Pages, 2020/08
Segawa, Tomoomi; Kawaguchi, Koichi; Ishii, Katsunori; Suzuki, Masahiro; Tachihara, Joji; Takato, Kiyoto; Okita, Takatoshi; Satone, Hiroshi*; Suzuki, Michitaka*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
To reduce the hold-up of the nuclear fuel materials in the glove box and the external exposure dose, the technology of the MOX powder adhesion prevention by the nanoparticle coating to the acrylic panels of the glove box has been developed. Due to the formation of nano-sized tiny rugged surface, the nanoparticle coating reduced the minimum adhesion force between the UO
particles and the acrylic test piece surface with the smallest particle size of about 5 
m where desorption was observed, by about one-tenth. Moreover, the nanoparticle coating reduced the amount of the MOX powder adhering to the acrylic test piece to about one-tenth. In this study, it was found that applying the nanoparticle coating to the acrylic panels of glove box can prevent the adhesion of nuclear fuel materials. This method is effective for reducing the hold-up of the nuclear fuel materials in the glove box, the external exposure dose and improving the visibility of the acrylic panels.
Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Itoi, Tatsuya*; Takada, Tsuyoshi*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 5 Pages, 2020/08
After the 2011 Fukushima accident, the seismic regulation for Nuclear Power Plants (NPP) have been strengthened to take countermeasures against accidents beyond design basis conditions. Therefore, the importance of seismic probabilistic risk assessment has drawn much attention. Uncertainty quantification is a very important issue in the fragility assessment for NPP buildings. In this study, the authors focus on the epistemic uncertainty that can be reduced, and aims to clarify the effects due to different modeling methods of NPP buildings on seismic response results. As the first step of this study, the authors compared the effects on seismic response using two kinds of modeling methods. In order to evaluate the effect, seismic response analysis was performed on two types of building models; the three dimensional finite element model and the conventional lumped mass with sway-rocking model. As the input ground motion, the authors adopted 200 types of simulated seismic ground motions generated by fault rupture models with stochastic seismic source characteristics. For the uncertainty quantification, the authors conducted statistical analyses of the effects on seismic response results of two kinds of modeling methods on building response for each input ground motions, and quantitatively evaluated the uncertainty of response considering different modeling methods. In particular, the difference in modeling methods clearly appeared near the openings of the floors and walls. The authors also report on the knowledge about these three-dimensional effects in seismic response analysis.
Okuda, Yukihiko; Kang, Z.; Nishida, Akemi; Tsubota, Haruji; Li, Y.
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 10 Pages, 2020/08
In case of projectile impact to reactor building of nuclear power plants, stress waves due to the projectile impact propagate from the impacted wall to the interior of the structure. It is an important issue to assess the dynamic response generated with projectile impact for safety related internal equipment because stress waves are likely to excite high-frequency vibrations of internal equipment in the reactor building. The OECD (Organization for Economic Co-operation and Development) / NEA (Nuclear Energy Agency) launched the IRIS (Improving Robustness Assessment Methodologies for Structures Impacted by Projectiles) benchmark project in order to assess the dynamic response for nuclear facility by projectile impact and the third phase of IRIS (IRIS 3) contributes to the investigation on the dynamic response of reinforced concrete (RC) structure with internal equipment. We have participated in the IRIS 3 and have performed the calibration analysis for projectile impact test on the structure which models a reactor building and internal equipment. Specially, we have developed and validated a numerical approach to investigate impact response of the RC structure with internal equipment through the calibration correction. This paper presents partial simulation results from dynamic response of the RC structure with internal equipment and discusses the effect of supporting condition of the internal equipment and stress wave propagation.
Nishida, Akemi; Kang, Z.; Okuda, Yukihiko; Tsubota, Haruji; Li, Y.
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 5 Pages, 2020/08
Studies on the local damage to reinforced concrete (RC) panels subjected to projectile impact have mainly focused on collisions that occur at an angle normal to the structure; thus, research on oblique impact is scarce. Therefore, we conducted research focusing on oblique impact to enable more realistic impact assessment of projectile collisions. To date, the validity of the analytical method has been confirmed by comparing the results with those of previous tests, and the local damage of RC panels that have collided with projectiles has been analytically investigated focusing on the impact angle. Therefore, this study aims to confirm the validity of the analysis method by conducting impact tests under various conditions including the impact angle, and obtaining data for validation. This paper outlines the test for the local damage of RC panels subjected to normal and oblique impact.
Taruta, Yasuyoshi; Yanagihara, Satoshi*; Hashimoto, Takashi*; Kobayashi, Shigeto*; Iguchi, Yukihiro; Kitamura, Koichi; Koda, Yuya; Tomoda, Koichi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 8 Pages, 2020/08
Decommissioning is a long-term project during which generations are expected to change. Therefore, it is necessary to appropriately transfer knowledge and technology to the next generation. In recent years, in the world of decommissioning, attempts have been made to apply advanced technologies such as utilization of knowledge management and virtual reality. This study describes adaptation in decommissioning from the viewpoint of utilizing IT technology called digital twin from the viewpoint of knowledge management.
Yamashita, Susumu; Kino, Chiaki*; Yoshida, Hiroyuki
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
In order to contribute the improvement of estimation accuracy for severe accident code such as SAMPSON, we have developed the chemical reaction model such as eutectic reaction and oxidation in micro scale, e.g., B
C-SUS in the control rod blade and UO-Zry in fuel rods, and implemented them to the computational fluid dynamics code named JUPITER. And we try to develop the coupled analysis frame work using SAMPSON and JUPITER to decrease uncertainty due to micro scale phenomena which cannot be calculate by severe accident analysis codes. From the preliminary analysis in fuel rod heating analysis by JUPITER using SAMPSON output data, it was revealed that the implemented chemical reaction models work stably and obtain reasonable results.
Trianti, N.; Motegi, Kosuke; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 9 Pages, 2020/08
Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro; Kikuchi, Shin; Emura, Yuki; Kamiyama, Kenji; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; et al.
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 10 Pages, 2020/08
One of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors is eutectic reactions between boron carbide (BC) and stainless steel (SS) as well as its relocation. Such behaviors have never been simulated in CDA numerical analyses in the past, therefore it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study focuses on BC-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in a range from solid to liquid state. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies conducted until 2018. Specific results in this paper are boron concentration distributions of solidified BC-SS eutectic sample in the eutectic melting experiments, which would be used for the validation of the eutectic physical model implemented into the computer code.
Sato, Hiroyuki; Aoki, Takeshi; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 10 Pages, 2020/08
The present study aims to propose a guidance that facilitates to determine fuel design limits of commercial HTGR on the basis of licensing experience through the HTTR construction. The guidance consists of a set of FOMs and a process to determine their evaluation criteria. The FOMs are firstly identified to satisfy safety requirements and a basic concept of safety guides established in a special committee under the AESJ with the support of the Research Association of High Temperature Gas Cooled Reactor Plant. The development process for the evaluation criteria takes into account not only the top-level regulatory criteria but also design dependent constraints including the performance of fission product containment in physical barriers other than fuel, fuel qualification criteria, design specifications of an instrumentation and control system. As a result, a comprehensive and transparent procedure for designers of prismatic-type commercial HTGR has been developed.
Yada, Hiroki; Takaya, Shigeru; Enuma, Yasuhiro
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
Aizawa, Kosuke; Hiyama, Tomoyuki; Nishimura, Masahiro; Kurihara, Akikazu; Ishida, Katsuji*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 8 Pages, 2020/08
A sodium-cooled fast reactor is designed to attain a high burn-up core in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, the deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the flow velocity distribution in a wire wrapped pin bundle. In this study, the detailed flow velocity distribution in the subchannel has been obtained by PIV (Particle Image Velocimetry) measurement using a wire-wrapped 3-pin bundle water model. Flow velocity conditions in the pin bundle were set from 0.036 m/s (
= 270) to 1.6m/s ( = 13,500). From the PIV results, the maximum flow velocity was increased by decreasing the number in the region away from the wrapping wire. Moreover, the PIV measurements by using the 3-pin bundle geometry without the wrapping wire were conducted. From the results, the effect of the wrapping wire on the flow field in the subchannel was understood. There experimental results useful not only for understanding of pin bundle thermal hydraulics but also code validation.
Igarashi, Kai*; Onuki, Ryoji*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
Kikuchi, Shin; Yamano, Hidemasa; Nakamura, Kinya*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 9 Pages, 2020/08
In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (BC) and stainless steel (SS) may take place. Thus, kinetic behavior of BC-SS eutectic melting is one of the important phenomena to be considered when evaluating the core disruptive accidents in SFR. In this study, for the first step to obtain the fundamental information on kinetic feature of BC-SS eutectic melting, the thermal analysis using the pellet type samples of BC and Type 316L SS as different experimental technique was performed. The differential thermal analysis endothermic peaks for the BC-SS eutectic melting appeared from 1483K to 1534K and systematically shifted to higher temperatures when increasing heating rate. Based on this kinetic feature, apparent activation energy and pre-exponential factor for the BC-SS eutectic melting were determined by Kissinger method. It was found that the kinetic parameters obtained by thermal analysis were comparable to the literature values.
Sun, Haomin; Leblois, Y.*; Gelain, T.*; Porcheron, E.*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 11 Pages, 2020/08
Aoki, Takeshi; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
The flow distribution analysis, which is a part of thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR) considering unintended flows between graphite blocks, has been performed for steady and conservative conditions. On the other hand, the transient analysis for satisfactorily realistic conditions will be helpful for the design improvement of prismatic-type HTGR. The present study aims to develop the transient flow distribution analysis code and confirm its applicability for the transient flow distribution analysis for prismatic-type HTGRs during anticipated operational occurrences and accidents utilizing experiences on high temperature engineering test reactor (HTTR) design. The calculation model and code were developed and validated for analysis of the unintended flows in the core and the molecular diffusion dominant in beginning air ingress behavior in an air ingress accident.
