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Miyahara, Shinya*; Arita, Yuji*; Nakano, Keita; Maekawa, Fujio; Sasa, Toshinobu; Obayashi, Hironari; Takei, Hayanori
Nuclear Engineering and Design, 403, p.112147_1 - 112147_17, 2023/03
Times Cited Count：0 Percentile：0.04(Nuclear Science & Technology)It is important to evaluate the inventories and the release and transport behavior of the spallation products (SPs) in the Lead-Bismuth Eutectic (LBE) coolant system of Accelerator Driven System (ADS) for the safety studies of the radiological hazard both in the cases of normal operation and accident. University of Fukui and JAEA have been developing the computer analysis code TRAIL (Transport of RAdionuclides In Liquid metal systems) which predicts the time dependent behavior of SPs within the LBE coolant system of ADS for the wide range of operational events. The source term of both radioactive and stable SPs in the LBE coolant is given as input and the radioactive decay chain model for the radioactive SPs is implemented in the code to evaluate the effect of precursors on the SPs mobility. This paper presents the recent advancement status of the code development and the validation results comparing with the distribution data of volatile SPs in MEGAPIE spallation target.
Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya*; Uno, Masayoshi*
Journal of Nuclear Science and Technology, 59(2), p.198 - 206, 2022/02
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Miyahara, Shinya*; Kawaguchi, Munemichi; Seino, Hiroshi; Atsumi, Takuto*; Uno, Masayoshi*
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 6 Pages, 2021/08
In a postulated accident of fuel pin failure of sodium cooled fast reactor, a fission product cesium will be released from the failed pin as an aerosol such as cesium iodide and/or cesium oxide together with a fission product noble gas such as xenon and krypton. As the result, the xenon and krypton released with cesium aerosol into the sodium coolant as bubbles have an influence on the removal of cesium aerosol by the sodium pool in a period of bubble rising to the pool surface. In this study, cesium aerosol removal behavior due to inertial deposition, sedimentation and diffusion from a noble gas bubble rising through liquid sodium pool was analyzed by a computer program which deals with the expansion and the deformation of the bubble together with the aerosol absorption considering the effects of particle size distribution and agglomeration in aerosols. In the analysis, initial bubble diameter, sodium pool depth and temperature, aerosol particle diameter and density, initial aerosol concentration in the bubble were changed as parameter, and the results for the sensitivities of these parameters on decontamination factor (DF) of cesium aerosol were compared with the results of the previous study in which the effects of particle size distribution and agglomeration in aerosols were not considered. From the results, it was concluded that the sensitivities of initial bubble diameter, the aerosol particle diameter and density to the DF became significant due to the inertial deposition of agglomerated aerosols. To validate these analysis results, the simulation experiments have been conducted using a simulant particles of cesium aerosol under the condition of room temperature in water pool and air bubble systems. The experimental results were compared with the analysis results calculated under the same condition.
studied by polarized and unpolarized neutron diffractionTerada, Noriki*; Qureshi, N.*; Stunault, A.*; Enderle, M.*; Ouladdiaf, B.*; Colin, C. V.*; Khalyavin, D. D.*; Manuel, P.*; Orlandi, F.*; Miyahara, Shin*; et al.
Physical Review B, 102(8), p.085131_1 - 085131_7, 2020/08
Times Cited Count：2 Percentile：16.85(Materials Science, Multidisciplinary)Miyahara, Shinya*; Kawaguchi, Munemichi; Seino, Hiroshi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
In a postulated accident of fuel pin failure of sodium cooled fast reactor, a fission product cesium will be released as an aerosol such as cesium iodide and/or oxide together with xenon and/or krypton. In this study, cesium aerosol removal behavior due to inertial deposition, sedimentation and diffusion was analyzed by a computer program which deals with the expansion and the deformation of the bubble together with the aerosol absorption. Initial bubble diameter, sodium pool depth and temperature, aerosol particle diameter and density, initial aerosol concentration were changed as parameter. From the results, it was concluded that the initial bubble diameter was most sensitive parameter to the decontamination factor (DF). It was found that the sodium pool depth, the aerosol particle diameter and density have also important effect on the DF, but the sodium temperature has a marginal effect. To meet these results, the experiments are under planning to validate the results.
Kawaguchi, Munemichi; Miyahara, Shinya*; Uno, Masayoshi*
Journal of Nuclear Engineering and Radiation Science, 6(2), p.021305_1 - 021305_9, 2020/04
Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel. In this study, SCR experiments with an internal heater were performed to investigate the chemical reaction beneath the internal heater (800
C), which was used to simulate the obstacle and heating effect on SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. The internal heater on the concrete hindered the transport of Na into the concrete. Therefore, Na could start to react with the concrete at the periphery of the internal heater, and the concrete ablation depth at the periphery was larger than under the internal heater. The high Na pool temperature of 800C increased largely the Na aerosol release rate, which was explained by Na evaporation and hydrogen bubbling, and formed the porous reaction product layer, whose porosity was 0.54-0.59 from the mass balance of Si and image analyzing EPMA mapping. They had good agreement with each other. The porous reaction products decreased the amount of Na transport into the reaction front. The Na concentration around the reaction front became about 30wt.% despite the position of the internal heater. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.
Miyahara, Shinya*; Ohdaira, Naoya*; Arita, Yuji*; Maekawa, Fujio; Matsuda, Hiroki; Sasa, Toshinobu; Meigo, Shinichiro
Nuclear Engineering and Design, 352, p.110192_1 - 110192_8, 2019/10
Times Cited Count：4 Percentile：47.64(Nuclear Science & Technology)Lead-Bismuth Eutectic (LBE) is used as a spallation neutron target and coolant materials of Accelerator Driven System (ADS), and many kinds of elements are produced as spallation products. It is important to evaluate the release and transport behavior of the spallation products in the LBE. The inventories and the physicochemical composition of the spallation products produced in LBE have been investigated for an LBE loop in the ADS Target Test Facility (TEF-T) in J-PARC. The inventories of the spallation products in the LBE were estimated using the PHITS code. The physicochemical composition of the spallation products in the LBE was calculated using the Thermo-Calc code under the conditions of the operation temperatures of LBE from 350C to 500C and the oxygen concentrations in LBE from 10 ppb to 1 ppm. The calculation showed that the 5 elements of Rb, Tl, Tc, Os, Ir, Pt, Au and Hg were soluble in LBE under the all given conditions and any kinds of compound were not formed in LBE. It was suggested that the oxides of Ce, Sr, Zr and Y were stable as CeO
, SrO, ZrO and YO in the LBE.
Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya*; Uno, Masayoshi*
Journal of Nuclear Science and Technology, 56(6), p.521 - 532, 2019/06
Times Cited Count：1 Percentile：13.41(Nuclear Science & Technology)Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*
Journal of Nuclear Science and Technology, 56(6), p.513 - 520, 2019/06
Times Cited Count：2 Percentile：25.74(Nuclear Science & Technology)This study revealed melting points and thermal conductivities of four samples generated by sodium-concrete reaction (SCR). We prepared the samples using two methods such as firing mixtures of sodium and grinded concrete powder, and sampling depositions after the SCR experiments. In the former, the mixing ratios were determined from the past experiment. The latter simulated the more realistic conditions such as the temperature history and the distribution of Na and concrete. The thermogravimetry-differential thermal analyzer (TG-DTA) measurement showed the melting points were 865-942C, but those of the samples containing metallic Na couldn't be clarified. In the two more realistic samples, the compression moldings in a furnace were observed. The observation revealed the softening temperature was 800-840C and the melting point was 840-850C, which was 10-20C lower than the TG-DTA results. The thermodynamics calculation of FactSage 7.2 revealed the temperature of the onset of melting was caused by melting of the some components such as NaSiO and/or Na
SiO. Moreover, the thermal conductivity was 
=1-3W/m-K, which was comparable to xNaO-1-xSiO (x=0.5, 0.33, 0.25), and those at 700C were explained by the equation of 
.
Horiguchi, Naoki; Miyahara, Naoya; Uesawa, Shinichiro; Yoshida, Hiroyuki; Osaka, Masahiko
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
For source term evaluation from reactor buildings (RBs) in LWR severe accidents, we have launched to develop an evaluation method of FP aerosol particle deposition onto surfaces of internal structures in an RB based on computational fluid dynamics (CFD). This paper describes development of a CFD simulation tool as the base part of the evaluation method. A preliminary simulation for a representative RB under a representative flow condition was conducted to confirm the tool performance by roughly grasping the deposition behaviors of FP aerosol particle and decontamination factor (DF) in the RB. Calculation results showed that most of aerosol particles were deposited along with gas flow formed by the internal structures in the RB, demonstrating the advantageous feature of the present CFD tool. The DFs from 4 to 14 were obtained with increase of the particle diameters from 0.1 to 10 
m as expected in terms of the particle movement equation.
Uesawa, Shinichiro; Miyahara, Naoya; Horiguchi, Naoki; Yoshida, Hiroyuki; Osaka, Masahiko
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05
Yoshida, Hiroyuki; Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Ose, Yasuo*
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11
Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*
Journal of Nuclear Science and Technology, 55(8), p.874 - 884, 2018/08
Times Cited Count：4 Percentile：42.98(Nuclear Science & Technology)As parts of severe accident studies in sodium-cooled fast reactor, experiments were performed to investigate the termination mechanism of sodium-concrete reaction (SCR). In the experiment, the reaction time was controlled to investigate the distribution change of sodium (Na) and the reaction products in the pool and around the reaction front. In the results, the Na around the reaction front decreased from the enough amount with the reaction time. The concentrations were 18-24 wt.% for Na, and 22-18 wt.% for Si after the termination. From the thermodynamics calculations, the stable materials around the reaction front comprised more than 90 wt.% solid products such as NaSiO, and no Na. Further, the distribution of Na and reaction products could be explained by a steady-state sedimentation-diffusion model. At the early stage of SCR, the reaction products were suspended as particles in the Na pool because of the high H-generation rate. As the concrete ablation proceeds, they start settling down due to the decreased H-generation rate, thereby allowing SCR termination. It was concluded that SCR termination was caused by the sediment of the reaction products and the lack of Na around the reaction front.
Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the presence of large sources of hydrogen and aerosols in the containment vessel. In this study, SCR experiments with an internal heater (800C) were performed to investigate the chemical reaction under the internal heater. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. Because the internal heater hindered the transport of Na, the moisture in the concrete, and reaction products, Na could permeate and react with the surface concrete at the periphery of the internal heater. As the SCR proceeded, the reaction products accumulated under the internal heater and disturbed the Na diffusion. Therefore, the Na concentration under the internal heater decreased relatively lower, and the concrete ablation depth under the internal heater decreased compared to that under the periphery of the internal heater. However, the Na concentration around the reaction front was about 30 wt.% despite the position of the internal heater. The Na concentration was similar to that of Na
SiO
, which was almost same as that in our past study. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.
Yoshida, Hiroyuki; Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Ose, Yasuo*
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
no abstracts in English
Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Yoshida, Hiroyuki
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
no abstracts in English
Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*
Netsu Sokutei, 45(1), p.2 - 8, 2018/01
Liquid sodium (Na) has been used as the coolant of fast reactors for the various merits, such as the high thermal conductivity. On the other hand, it is postulated that a steel liner may fail and lead to a sodium-concrete reaction (SCR) during the Na-leak accident. Because of concrete ablation and release of hydrogen gas due to the chemical reactions between Na and concrete components, the SCR is one of the important phenomena in the Na-leak accident. In the study, fundamental experiments related to the SCR were performed using Na and concrete powder. Here, the used concrete powder is milled siliceous concrete which is usually used as the structural concrete in Japanese nuclear power plants. The obvious temperature changes at 3 temperature regions were observed for the reaction process such as Na-melt, NaOH-SiO and Na-HO-SiO reaction, which occurred around 100, 300 and 500C, respectively. Especially, the violent reaction around 500C caused the temperature peak to 
C, and the reaction heat of 
kW/g was estimated under the Na-concrete mixing ratio such as 
. The main components of the reaction products was identified as NaSiO with X-ray diffraction technique. Moreover, the measured thermophysical properties such as melting point, density, specific heat, thermal conductivity and viscosity were similar to those of 
NaO-
SiO (
).
Kawaguchi, Munemichi; Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya
Journal of Nuclear Science and Technology, 53(12), p.2098 - 2107, 2016/12
Times Cited Count：6 Percentile：52.61(Nuclear Science & Technology)A sodium concrete reaction (SCR) is one of the important phenomena to cause the structural concrete ablation and the release of H gas in the case of sever accident of sodium cooled fast reactors. In this study, the long-time SCR test had been carried out to investigate the self-termination mechanism. The results showed the SCR terminated even if the enough amount of Na remained on the concrete. The quantitative data were collected on the SCR terminating such as temperature and H generation. The reaction products, which became the small solids in liquid Na were transferred with slurry state by generated H bubbles. Though the Na transfers actively and ablated the concrete surface with the high H generation rate, the mass exchange coefficient defined as 
decreased and the reaction products settled gradually with decreasing the H generation rate. Therefore, the Na concentration decreased at the reaction front and resulted in the SCR terminating naturally.
Kawaguchi, Munemichi; Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
CONTAIN-LMR code is an integrated analysis tool to predict the consequence of severe accident in a liquid metal fast reactor. A sodium-concrete reaction is one of the most important phenomena, and Sodium-Limestone Concrete Ablation Model (SLAM) has been installed into the original CONTAIN code. The SLAM treats chemical reaction kinetics between the sodium and the concrete compositions mechanistically, the application is limited to the limestone concrete. In order to apply SLAM to the siliceous concrete which is an ordinary structural concrete in Japan, the chemical reaction kinetics model has been improved. The improved model was validated to analyze a series of sodium-concrete experiments which were conducted in Japan Atomic Energy Agency. It has been found that relatively good agreement between calculation and experimental results is obtained and the CONTAIN-LMR code has been validated with regard to the sodium-concrete reaction phenomena.
Miyahara, Shinya; Seino, Hiroshi; Ohno, Shuji; Konishi, Kensuke
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
A CONTAIN-LMR code has been developed in JAEA for application to PRA of LMFRs since the original CONTAIN code had been introduced from SNL of U.S. in 1982. The code is a best-estimate, integrated analysis tool for predicting the physical, chemical and radiological conditions inside a containment building of LMFRs following a severe accident with reactor vessel melt-through. The code is also able to predict the source term to the environment in the accident. This code can treat many important phenomena consistently such as sodium fire, radioactive aerosol behavior, hydrogen burn, sodium-concrete reaction and core debris-concrete interaction occurred in the accident with inter-cell heat and mass flow under the multiple cell geometry. This paper describes the chronology of the code development in JAEA briefly as an introduction, and after that, the outline of computational models in the code, the examples of the code validation, and the future plan of the code application are described.
