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Journal Articles

Leaching behavior of prototypical Corium samples; A Step to understand the interactions between the fuel debris and water at the Fukushima Daiichi reactors

Nakayoshi, Akira; Jegou, C.*; De Windt, L.*; Perrin, S.*; Washiya, Tadahiro

Nuclear Engineering and Design, 360, p.110522_1 - 110522_18, 2020/04

 Times Cited Count:1 Percentile:100(Nuclear Science & Technology)

JAEA Reports

The In-situ experiment for performance confirmation of engineered barrier system at Horonobe Underground Research Laboratory; Installation of engineered barrier system and backfilling the test niche at the 350m gallery

Nakayama, Masashi; Ono, Hirokazu

JAEA-Research 2019-007, 132 Pages, 2019/12

JAEA-Research-2019-007.pdf:11.29MB
JAEA-Research-2019-007-appendix(CD-ROM).zip:39.18MB

The Horonobe Underground Research Laboratory (URL) Project has being pursued by Japan Atomic Energy Agency (JAEA) to enhance the reliability of relevant disposal technologies through investigations of the deep geological environment within the host sedimentary formation at Horonobe, northern Hokkaido. The URL Project consists of two major research areas, "Geoscientific Research" and "Research and Development on Geological Disposal Technologies". The in-situ experiment for performance confirmation of engineered barrier system (EBS experiment) had been prepared from 2013 to 2014 fiscal year at GL-350m gallery, and heating by electric heater in simulated overpack had started in January, 2015. One of objectives of the EBS experiment is acquiring data concerned with Thermal-Hydrological-Mechanical-Chemical (THMC) coupled behavior. These data will be used in order to confirm the performance of engineered barrier system. This report shows following works had carried out at the GL-350 m gallery. Excavation of a test niche and a test pit, Setting buffer material blocks and a simulated overpack into the test pit, Backfilling the niche by compaction backfilling material and setting backfilling material blocks, Casting concrete type plug and contact grouting, Consolidate measurement system and start measuring.

Journal Articles

Diffusion and sorption behavior of HTO, Cs, I and U in mortar

Akagi, Yosuke*; Kato, Hiroyasu*; Tachi, Yukio; Sakamoto, Hiroyuki*

Progress in Nuclear Science and Technology (Internet), 5, p.233 - 236, 2018/11

A large amount of radioactive contaminated concrete will be generated from the decommissioning in the Fukushima Dai-ichi Nuclear Power Plant (NPP). For developing the plans of decommissioning and waste management including decontamination and disposal, it is important to estimate radionuclides inventory and concentration distribution in the concrete materials. In this study, effective diffusivities (De) and distribution coefficients (Kd) of HTO, Cs, I and U in OPC mortar were measured by through-diffusion and batch sorption experiments. De values derived were in the sequence of HTO, I, Cs, U, implying that cation exclusion effects may be important mechanisms in OPC mortar. Kd values derived by batch tests were higher by more than one order of magnitude than the diffusion-derived Kd values, indicating that crushing of samples had a strong influence on sorption. Diffusion and sorption mechanisms in OPC mortar were evaluated to predict the penetration behavior of these radionuclides.

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:3 Percentile:38.72(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

Three-dimensional numerical study on pool stratification behavior in molten corium-concrete interaction (MCCI) with MPS method

Li, X.; Sato, Ikken; Yamaji, Akifumi*; Duan, G.*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.

Journal Articles

Deformation analysis of reinforced concrete using neutron imaging technique

Koyama, Taku*; Ueno, Kazuki*; Sekine, Mariko*; Matsumoto, Yoshihiro*; Kai, Tetsuya; Shinohara, Takenao; Iikura, Hiroshi; Suzuki, Hiroshi; Kanematsu, Manabu*

Materials Research Proceedings, Vol.4, p.155 - 160, 2018/05

 Times Cited Count:0 Percentile:100

JAEA Reports

The In-situ experiment for performance confirmation of engineered barrier system at Horonobe Underground Research Laboratory; Consideration of concrete-type plug composition

Nakayama, Masashi; Niunoya, Sumio*; Miura, Norihiko*; Takeda, Nobufumi*

JAEA-Research 2017-016, 62 Pages, 2018/01

JAEA-Research-2017-016.pdf:19.99MB

The Horonobe URL Project has being pursued by JAEA to enhance the reliability of relevant disposal technologies through investigations of the deep geological environment within the host sedimentary formation at Horonobe, Hokkaido. The URL Project consists of 2 major research areas, "Geoscientific Research" and "Research and Development on Geological Disposal Technologies", and proceeds in 3 overlapping phases, over a period of around 20 years. Phase III (Investigations in the underground facilities) investigation was started in 2010 FY. The in-situ experiment for performance confirmation of engineered barrier system (EBS experiment) had started 2013 at GL-350 m gallery, and heating by electric heater in simulated overpack had started in January, 2015. One of objectives of the EBS experiment is acquiring data concerned with THMC coupled behavior. These data will be used in order to confirm the performance of engineered barrier system. This report shows consideration of concrete-type plug composition. The low alkaline cement developed by JAEA, called HFSC, was used for the plug. HFSC has used in construction test at the gallery as shotcrete in Horonobe URL.

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

Monte Carlo criticality analysis under material distribution uncertainty

Ueki, Taro

Journal of Nuclear Science and Technology, 54(3), p.267 - 279, 2017/03

 Times Cited Count:4 Percentile:38.75(Nuclear Science & Technology)

Analysis framework under material distribution uncertainty is investigated for the Monte Carlo (MC) criticality calculation of continuously mixed media formed via molten core concrete interaction. Deterministic trigonometric functions and randomized Weierstrass functions are utilized to represent the spatially continuous variation. Numerical results indicate that the effective multiplication factor (k$$_{rm eff}$$) under random spatial variation can depart significantly from the k$$_{rm eff}$$ of a reference uniform medium. It is also shown that the deterministic modeling provides an upper-bound measure for extreme results from random realizations.

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:4 Percentile:46.36(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.

JAEA Reports

Study of treatment method for damaged fuel removed from the spent fuel pool; Outline of annual report for JFY 2013 and 2014 (Contract research)

Iijima, Shizuka; Uchida, Naoki; Taguchi, Katsuya; Washiya, Tadahiro

JAEA-Review 2015-018, 39 Pages, 2015/11

JAEA-Review-2015-018.pdf:3.95MB

There is a possibility that the fuel assemblies stored in the spent fuel pool (SFP) at Fukushima Daiichi NPS (or Nuclear Power Station) are not only exposed to seawater and concrete fragments, but also damaged by fallen rubbles. We checked the reprocessing experiences of leak fuels at Tokai Reprocessing Plant and overseas reprocessing facilities, and the storage conditions and the checked and inspected results of the fuel stored in the SFP at Fukushima Daiichi NPS, after that, we listed up the technological problems with reprocessing damaged nuclear fuels and selected elements of the research for the purpose of drawing indicators to make a judgmental decision of the possibility of damaged nuclear fuels reprocessing. And we drew the indicators to make a judgmental decision on the possibility of reprocessing based on the results of the examination and the study about elements of the research.

Journal Articles

Analyses of iron and concrete shielding experiments at JAEA/TIARA with JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0

Konno, Chikara; Ochiai, Kentaro; Sato, Satoshi; Ota, Masayuki

Fusion Engineering and Design, 98-99, p.2178 - 2181, 2015/10

 Times Cited Count:4 Percentile:53.79(Nuclear Science & Technology)

We have analyzed the iron and concrete shielding experiments with the 40 and 65 MeV neutron sources at TIARA in Japan Atomic Energy Agency with the latest high-energy nuclear data libraries, JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0. The Monte Carlo code MCNP-5 and ACE files of JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0, which were supplied from JAEA, BNL and IAEA, respectively, were used for this analysis. The followings are found out from the results. (1) The calculations with JENDL/HE-2007 agree with all the measured ones well; (2) Those with ENDF/B-VII.1 tend to overestimate the measured ones with the thickness of the assemblies largely; (3) Those with FENDL-3.0 agree with the measured ones well for the iron experiment, while they overestimate the measured ones well for the concrete experiment largely. Some data in ENDF/B-VII.1 and FENDL-3.0 should be revised.

Journal Articles

Application of neutron stress measurement to reinforced concrete structure

Suzuki, Hiroshi; Kusunoki, Koichi*; Kanematsu, Manabu*; Tasai, Akira*; Hatanaka, Yuichi*; Tsuchiya, Naoko*; Bae, S.*; Shiroishi, Sho*; Sakurai, Sonoko*; Kawasaki, Takuro; et al.

JPS Conference Proceedings (Internet), 8, p.031006_1 - 031006_6, 2015/09

The bond resistance between reinforcing bar (rebar) and concrete is one important parameter for examining integrity of the reinforced concrete structure, and is commonly evaluated by measuring the strain distribution along the rebar embedded in concrete. Here we present two types of applications of the time-of-flight neutron diffraction technique to measure the stress distribution of the rebar. It was demonstrated that bond deterioration around cracks developed in concrete can be evaluated by measuring the stress distribution along the embedded rebar using neutron diffraction. Furthermore, a change in the stress distribution along the rebar due to bond deterioration by corrosion was observed. The neutron diffraction technique is expected to become a novel method for measuring the stress (strain) of the rebar embedded in concrete, and will eventually provide insight into the actual phenomena on the reinforced concrete structures.

Journal Articles

Influence of contaminants from spent fuel pools at the Fukushima Daiichi Nuclear Power Station on the reprocessing process

Aihara, Haruka; Kitawaki, Shinichi; Nomura, Kazunori; Taguchi, Katsuya

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1076 - 1083, 2015/09

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, 5; Improvement of debris-concrete interaction model

Seino, Hiroshi; Kawaguchi, Munemichi; Izumi, Keitaro*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

As a part of development of CONTAIN-LMR, CORCON and VANESA models for calculating the debris-concrete interaction (MCCI) have been improved taking into account the influence of soudium-pool existence. In this study, the following LMFR specific models in the code have been developed and improved: (1) chemical reaction in sodium pool, (2) aerosol decontamination in sodium pool, and (3) heat conduction in concrete. These models have been also confirmed and validated with experimental results. As a result, improved CORCON and VANESA can represent the MCCI behavior reasonably well. Further improvement and validation of CONTAIN-LMR will be continued in order to apply to the ex-vessel accident progression of LMFRs.

Journal Articles

In-situ dismantling of the liquid waste storage tank in the decommissioning program of the JRTF

Mimura, Ryuji; Muraguchi, Yoshinori; Nakashio, Nobuyuki; Nemoto, Koichi; Shiraishi, Kunio

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

The JAERI's Reprocessing Test Facility (JRTF) was the first engineering-scale reprocessing facility constructed in Japan. The JRTF was operated from 1968 to 1969 to reprocess spent fuels from the Japan Research Reactor No.3 (JRR-3). As a result of the operation (total 3 runs) by PUREX process, 200 g of highly purified plutonium (Pu) were extracted. In this operation, about 70 m$$^{3}$$ of liquid waste was generated and part of this waste, which including Pu, with relatively high radioactivity, was stored in six large tanks. After shutdown of the facility, the JRTF decommissioning program was started in 1990 to develop decommissioning technologies and to obtain experiences and data on dismantling of fuel cycle facilities. Liquid waste in the tanks was treated from 1982 to 1998. Dismantling of tanks started in 2002. The tanks were installed in narrow concrete cells and inside of the cell was high dose area. Dismantling method for the tank is important factor to decide manpower and time for dismantlement. In this paper, in-situ dismantling of the liquid waste storage tank and its preparation work are discussed.

Journal Articles

A Study of mechanical stability of support elements and surrounding rock mass during shaft sinking through a fault

Tsusaka, Kimikazu*; Inagaki, Daisuke*; Nago, Makito*; Ijiri, Yuji*

Proceedings of 8th Asian Rock Mechanics Symposium (ARMS-8) (USB Flash Drive), 9 Pages, 2014/10

Journal Articles

Reduction of contaminated concrete waste by recycling aggregate with the aid of pulsed power discharge

Arifi, E.*; Ishimatsu, Koichi*; Iizasa, Shinya*; Namihira, Takao*; Sakamoto, Hiroyuki*; Tachi, Yukio; Kato, Hiroyasu*; Shigeishi, Mitsuhiro*

Construction and Building Materials, 67(Part B), p.192 - 196, 2014/09

 Times Cited Count:2 Percentile:79.42(Construction & Building Technology)

The Fukushima Dai-ichi Nuclear Plant accident has resulted in a large amount of radioactively contaminated concrete. The possible application of the pulsed power discharge to reduce the amount of contaminated concrete as radioactive waste was investigated. The contaminated concrete was decontaminated by separating contaminated matrix from uncontaminated coarse aggregate under pulsed power discharge process. In this study, a stable Cs isotope was used to simulate radioactively contaminated concrete. As a result, while the volume of reclaimed aggregate from contaminated concrete could be reproduced was up to 60%, nevertheless Cs detected in the reclaimed aggregate was only approximately 3%. Thus most of the Cs were dissolved in water during the discharge process. It is expected that the pulsed power could reduce the contaminated concrete waste by reusing aggregate. Further investigations are requested to test the applicability of this method under the realistic conditions close to the actual waste.

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