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Journal Articles

Visualizing an ignition process of hydrogen jets containing sodium mist by high-speed imaging

Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya*; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 56(6), p.521 - 532, 2019/06

Journal Articles

Melting behavior and thermal conductivity of solid sodium-concrete reaction product

Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 56(6), p.513 - 520, 2019/06

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 865-942$$^{circ}$$C, 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 800-840$$^{circ}$$C and the melting point was 840-850$$^{circ}$$C, which was 10-20$$^{circ}$$C 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 Na$$_{2}$$SiO$$_{3}$$ and/or Na$$_{4}$$SiO$$_{4}$$. Moreover, the thermal conductivity was $$lambda$$=1-3W/m-K, which was comparable to xNa$$_{2}$$O-1-xSiO$$_{2}$$ (x=0.5, 0.33, 0.25), and those at 700$$^{circ}$$C were explained by the equation of $$NBO/T$$.

Journal Articles

A Study on self-terminating behavior of sodium-concrete reaction, 2

Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 55(8), p.874 - 884, 2018/08

 Times Cited Count:2 Percentile:16.17(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 Na$$_{2}$$SiO$$_{3}$$, 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$$_{2}$$-generation rate. As the concrete ablation proceeds, they start settling down due to the decreased H$$_{2}$$-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.

Journal Articles

Discussion about sodium-concrete reaction in presence of internal heater

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 (800$$^{circ}$$C) 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$$_2$$SiO$$_3$$, 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.

Journal Articles

Thermophysical properties of sodium-concrete reaction products

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$$_{2}$$ and Na-H$$_{2}$$O-SiO$$_{2}$$ reaction, which occurred around 100, 300 and 500$$^{circ}$$C, respectively. Especially, the violent reaction around 500$$^{circ}$$C caused the temperature peak to $$836 sim 853^{circ}$$C, and the reaction heat of $$0.15 sim 0.23$$ kW/g was estimated under the Na-concrete mixing ratio such as $$gammaapprox 0.32$$. The main components of the reaction products was identified as Na$$_{2}$$SiO$$_{3}$$ 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 $$x$$Na$$_{2}$$O-$$(1-x)$$SiO$$_{2}$$ ($$xleq 0.5$$).

Journal Articles

A Study on self-terminating behavior of sodium-concrete reaction

Kawaguchi, Munemichi; Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya

Journal of Nuclear Science and Technology, 53(12), p.2098 - 2107, 2016/12

 Times Cited Count:3 Percentile:42.29(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$$_2$$ 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$$_2$$ generation. The reaction products, which became the small solids in liquid Na were transferred with slurry state by generated H$$_2$$ bubbles. Though the Na transfers actively and ablated the concrete surface with the high H$$_2$$ generation rate, the mass exchange coefficient defined as $$E_p$$ decreased and the reaction products settled gradually with decreasing the H$$_2$$ generation rate. Therefore, the Na concentration decreased at the reaction front and resulted in the SCR terminating naturally.

Journal Articles

Development of fast reactor containment safety analysis code, CONTAIN-LMR, 3; Improvement of sodium-concrete reaction model

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.

Journal Articles

Development of fast reactor containment safety analysis code, CONTAIN-LMR, 1; Outline of development project

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.

Journal Articles

Combustion characteristics of generating hydrogen during sodium-concrete reaction

Seino, Hiroshi; Ohno, Shuji; Yamamoto, Ikuo*; Miyahara, Shinya

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12

A hydrogen combustion experiment was conducted to simulate the sodium-concrete reaction under oxygen-existing conditions. As a result, it was found that hydrogen was burnt at the sodium pool surface because as sodium combustion heat played a role of the ignition energy, and the hydrogen combination ratio increased with the increase of the oxygen concentration in the atmosphere.

Journal Articles

Simulation of radioactive corrosion product in primary cooling system of Japanese sodium-cooled fast breeder reactor

Matsuo, Yoichiro; Miyahara, Shinya; Izumi, Yoshinobu*

Journal of Power and Energy Systems (Internet), 6(1), p.6 - 17, 2012/03

Radioactive corrosion product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE code. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE code incorporating the Particle Model.

JAEA Reports

Sodium fire test at broad ranges of temperature and oxygen concentrations, 4; Low temperature sodium spray tests (Translated document)

Kawata, Koji; Matsuki, Takuo*; Miyahara, Shinya

JAEA-Review 2011-046, 42 Pages, 2012/02

JAEA-Review-2011-046.pdf:2.41MB

Sodium spray fire tests at an initial sodium temperature of 250$$^{circ}$$C were conducted under the atmospheric conditions of air and 3% oxygen containing nitrogen to determine both the sodium burning rate and the aerosol release fraction.

Journal Articles

R&D on maintenance technologies for FBR plants in JAEA; The Status quo and the future plan

Tsukimori, Kazuyuki; Ueda, Masashi; Miyahara, Shinya; Yamashita, Takuya

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00

Journal Articles

Predictive analysis of the radiation exposure for the primary cooling system of the rated power operation of MONJU

Matsuo, Yoichiro; Hasegawa, Masanori; Maegawa, Yoshiharu; Miyahara, Shinya

Hoken Butsuri, 46(4), p.304 - 313, 2011/12

Radioactive corrosion products (CP) are main source of personal radiation exposure during maintenance without fuel-failure accident in the liquid-metal fast breeder reactor (LMFBR) plants. In order to establish the techniques of radiation dose estimation for personnel, program system DORE has been developed. The DORE system is constructed by PSYCHE code and QAD code system. The density of each deposited CP of primary coolant system in MONJU was estimated by using the PSYCHE. Moreover, the QAD-CGGP2R code is applied to dose rate calculations for the primary coolant system in MONJU. The predicted values were estimated to be saturated at 2-3 mSv/h in twenty years after the start of operation, and the dose rate reaches 4 mSv/h in domains near the IHX and the cold-leg piping.

Journal Articles

The Dependence of equilibrium partition coefficient of cesium and iodine between sodium pool and the inert cover gas on the concentration in the pool

Miyahara, Shinya; Nishimura, Masahiro; Nakagiri, Toshio

Nuclear Engineering and Design, 241(12), p.4731 - 4736, 2011/12

 Times Cited Count:2 Percentile:75.97(Nuclear Science & Technology)

We measured equilibrium partition coefficients of cesium and iodine between liquid sodium pool and the inert cover gas. The obtained empirical equations were consistent with Castleman's theoretical equations. The effect of cesium concentration upon the partition coefficients was consistent with the theoretical values. That of iodine concentration was incompatible with the theoretical consideration due to the formation of dimmer of NaI (Na$$_{2}$$I$$_{2}$$) in the cover gas.

Journal Articles

Experimental investigation of reaction behavior between carbon dioxide and liquid sodium

Miyahara, Shinya; Ishikawa, Hiroyasu; Yoshizawa, Yoshio*

Nuclear Engineering and Design, 241(5), p.1319 - 1328, 2011/05

 Times Cited Count:7 Percentile:39.29(Nuclear Science & Technology)

Reaction behavior of carbon dioxide (CO$$_{2}$$) with a liquid sodium pool was experimentally investigated to understand the consequences of boundary tube failure in a sodium-CO$$_{2}$$ heat exchanger (HX). In this study, two kinds of experiments, namely fundamental experiment and demonstration experiment which simulate the incident of CO$$_{2}$$ leakage in HX, were carried out to investigate the reaction behavior. From these experiments, it became clear that the exothermic reaction occurred above a threshold temperature, and useful and indispensable information such as the resulting temperature and pressure rise and the behavior of solid reaction products in the pool was obtained to evaluate the consequences of boundary tube failure incident in the sodium- CO$$_{2}$$ heat exchanger.

Journal Articles

Proposed method of the modeling and simulation of corrosion product behavior in the primary cooling system of fast breeder reactors

Matsuo, Yoichiro; Miyahara, Shinya; Izumi, Yoshinobu*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/05

In order to establish the techniques of radiation dose estimation for worker in radiation-controlled area, PSYCHE code has been developed. The PSYCHE is based on the Solution-Precipitation model. The CP transfer calculation using the Solution-Precipitation model needs a fitting factor for the calculation of the precipitation of CP. In this study, in addition to existing Solution-Precipitation model in PSYCHE, a transfer-model of CP species in particle form was applied to calculations of CP behavior in the primary cooling system of fast breeder reactor MONJU. Based on the calculated results, we estimated the contribution of CP deposition in the particle-form. It was suggested that the improved model including transfer-model of CP species in particle-form could be used for evaluation of CP transfer and radiation-source distribution in place of conventional Solution-Precipitation model with fitting factor in the PSYCHE.

Journal Articles

Reactive wetting of metallic plated steels by liquid sodium

Kawaguchi, Munemichi; Tagawa, Akihiro; Miyahara, Shinya

Journal of Nuclear Science and Technology, 48(4), p.499 - 503, 2011/04

The sodium wetting experiments were performed to investigate the reactive wetting of metallic plating materials by liquid sodium at 250 $$^{circ}$$C for the ultrasonic sensor of the under-sodium viewer. SUS304 stainless steel specimens were electrolytically plated with four metallic materials (Nickel, Palladium, Gold and Indium) which have different solubility in sodium, and the spreading velocity of sodium droplets on the metallic plated specimens was measured. It was confirmed that the spreading velocity increased as the solubility increased, and the constant $$alpha$$ on the spreading velocity on the plated specimens was unique for the plating materials and was proportional to the logarithm of the solubility of the plating materials. Furthermore, it is considered possible to select plating materials based on solubility from the result of this study.

Journal Articles

Transport of radioactive corrosion products in primary system of sodium-cooled fast breeder reactor "MONJU"

Matsuo, Yoichiro; Hasegawa, Masanori; Maegawa, Yoshiharu; Miyahara, Shinya

Journal of Power and Energy Systems (Internet), 5(1), p.96 - 107, 2011/01

Radioactive corrosion products (CP) are primary cause of personal radiation exposure during maintenance work at FBR plants with no breached fuel. The PSYCHE code has been developed based on the solution-precipitation model for analysis of CP transfer behavior. We predicted and analyzed the CP solution and precipitation behavior of MONJU to evaluate the applicability of the PSYCHE code to MONJU, using the parameters verified in the calculations for JOYO. From the calculation result pertaining to the MONJU system, distribution of $$^{54}$$Mn deposited in the primary cooling system over 20 years of operation is predicted to be approximately 7 times larger than that of $$^{60}$$Co. In particular, predictions show a notable tendency for $$^{54}$$Mn precipitation to be distributed in the primary pump and cold-leg. The calculated distribution of $$^{54}$$Mn and $$^{60}$$Co in the primary cooling system of MONJU agreed with tendencies of measured distribution of JOYO.

Journal Articles

Development of level 2 PSA methodology for sodium-cooled fast reactors, 6; Development of technical basis in ex-vessel accident sequences

Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya

Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10

This research has compiled technical basis which is necessary to carry out a probabilistic safety assessment (Level 2 PSA) for a sodium-cooled fast reactor. The accumulated technical information consists of experimental and analytical information which help ones to understand the loading to a containment vessel, as well as the existing information on dominant factors of important ex-vessel phenomena.

Journal Articles

Prediction of radioactive corrosion product transfer in primary systems of Japanese prototype fast breeder reactor Monju

Matsuo, Yoichiro; Hasegawa, Masanori; Maegawa, Yoshiharu; Miyahara, Shinya

Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 8 Pages, 2010/05

Radioactive corrosion products are main cause of personal radiation exposure during maintenance with no breached fuel in FBR plants. CP is produced in the core region by activation of fuel cladding and sub-assembly wrappers, and they are transported to the primary circuit with sodium flow and deposited on the wall of the primary piping and components. In order to establish the techniques of radiation dose reduction for of personnel, program system for corrosion hazard evaluation code PSYCHE has been was developed. The PSYCHE code is based on the solution-precipitation model. The density of each deposited CP and dose rate of primary coolant system in Monju was estimated by using the PSYCHE and QAD-CG code.

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