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Motoshima, Takayuki*; Matsui, Hiroya; Kawakubo, Masahiro*; Kobayashi, Masato*; Ichimura, Tetsuhiro*; Sugita, Yutaka
Nihon Genshiryoku Gakkai-Shi ATOMO
, 64(3), p.163 - 167, 2022/03
no abstracts in English
Yabana, Shuichi*; Matsuda, Akihiro*; Nakahira, Masataka; Osaki, Toshio*
Nihon Kenchiku Gakkai 2002-Nendo Taikai Koen Kogaishu, P. 495, 2002/08
no abstracts in English
Isogai, Takeshi*; Oda, Chie
JNC TN8400 2000-025, 48 Pages, 2000/09
Porewater chemistly in compacted bentonite would affect a performance of engineered barrier system in a high-level radioactive waste repository, whereas there are little information of the porewater based on experimental data. The previous study provided a new method of direct pH measurement for highly compacted bentonite system and demonstrated some tests for compacted bentonite samples (the dry densities: 1.6 [g/cm
] and 1.8 [g/cm]) both with the de-ionized water and with the NaCl solution. In this study, the solution equilibrated with low alkalinity cement were used in the direct pH measurement to see the effect of the composition of the external solutions, in which the bentonite column immersed. The result showed that the pH value of porewater in the cementitious condition was around 9 during the immersed time 1 to 3 months, while after 6 months became the porewater pH 10.6, which was equal to pH of the external solution.
Furuichi, Mitsuaki*; Toida, Masaru*; Masumoto, Kazuhiko*
JNC TJ8400 2000-021, 196 Pages, 2000/02
JNC-TJ8400-2000-021.pdf:23.23MB
For the geological disposal of high level radioactive wastes, after placement of tbe wastes, it is necessary to close off (to be called "sealing" hereafter) the underground potential passages (disposal pits, disposal tunnels, main and connecting tunnels and access tunnels) with an effective combination of engineered barriers such as buffers, backfilling materials, plugs and grout. It is necessary to ensure the long-term durability to isolate disposed wastes in the system. The results of the research works this year are as follows; (1)The objectives are to discuss the design of tunnel sealing experiments at URL site. The results of research were about (a)tracer experiment and numerical analysis (b)evaporation measurement (c)presentation at the coordination meeting (2)The discussion was about the equipment of inclined compaction methods and bearing capacity of rock against pressures for the concrete plugs.
Toida, Masaru*; Masumoto, Kazuhiko*; *; Okutsu, Kazuo*; *
JNC TJ8400 2000-020, 68 Pages, 2000/02
JNC-TJ8400-2000-020.pdf:9.45MB
For the geological disposal of high level radioactive wastes, after placement of the wastes, it is necessary to close off (to be called "sealing" hereafter) the underground potential passages (disposal pits, disposal tunnels, main and connecting tunnels and access tunnels) with an effective combination of engineered barriers such as buffers, backfiling materials, plugs and grout. It is necessary to ensure the long-term durability to isolate disposed wastes in the system. The results of the research works this year are as follows ; (1)The objectives are to discuss the design of tunnel sealing experiments at URL site. The results of research were about (a)tracer experiment and numerical analysis (b)evaporation measurement (c)presentation at the coordination meeting. (2)The discussion was about the equipment of inclined compaction methods and bearing capacity of rock against pressures for the concrete plugs.
Sugita, Yutaka; Fujita, Tomoo; Tanai, Kenji; Hasegawa, Hiroshi; Furuichi, Mitsuaki*; Okutsu, Kazuo*; Miura, K.*
JNC TN8400 99-039, 58 Pages, 1999/11
Regarding disposal techniques of high-level radioactive waste (HLW), the HLW is vitrified and then stored for cooling for a period of 30 to 50 years. After cooling, the HLW is isolated in the deep underground. The concept of geological disposal is based on the requirement to enclose the HLW in the deep underground for the long-term durability of the human's environmental safety. Backfilling of a repository is a unique activity on the geological disposal. If underground tunnels excavated to construct the repository are left, they may have significant influences on the barrier performance of an entire repository, such as: the mechanical stability of a tunnel may be damaged by rock stresses and a tunnel may provide a fast pathway for ground water flow. Therefore, the underground facilities are expected to be backfilled with a backfilling material after emplacement of the HLW and a buffer material. The material for the backfilling of the underground facilities is backfilling material. In this report, bentonite-aggregate mixture is considered, as one of the candidate materials for the backfilling material. Aggregate imitates the muck that is generated during construction phase of the underground facilities. The combination of backfilling, plugging and grouting is considered in some underground situations. Plug is composed of concrete material or clay-based one. Grouting material is concrete material or clay-based one, too. In this report, the concept of the backfilling, mechanical and hydrological characteristics of the bentonite-aggregate mixture, the function, work methods and a schedule of the backfilling materials, plugging and grouting are considered, and items of quality control for the bentonite-aggregate mixture, concrete material and grouting are listed.
Sato, Satoshi; Takatsu, Hideyuki; *; Iida, Hiromasa; Mori, Seiji*; R.Santoro*
Fusion Engineering and Design, 42, p.213 - 219, 1998/00
Times Cited Count：1 Percentile：15.03(Nuclear Science & Technology)no abstracts in English
Hibiya, Keisuke*; *; Shiogama, Yukihiro*; Masumoto, Kazuhiko*; Fukazawa, E.*; Taira, K.*; Tanaka, Toshiyuki*; Kondo, Y.*; Yamamoto, M.*; Okutsu, Kazuo*; et al.
PNC TJ1100 97-004, 69 Pages, 1997/02
None
*;
PNC TN9410 96-102, 40 Pages, 1996/04
In designing the shield plug of LMFBR, it is important to evaluate the thermal response between the cover gas thermal-hydraulics and the temperature fields of the shield plug at the same time. Based on the experiments which were performed by OEC, the natural convection and the thermal radiation in the cover gas layer were calculated with the structure simulating the shield plug in a detail two-dimensional model. The calculations were carried out for 8 kinds of experimental RUNs using a FLUSH code. The main results were as follows: (1)For these 8 kinds of experimental RUNs, the velocity and the temperature distributions in the cover gas layer were presented. The radial and axial temperature distributions in the rotating plug were also presented, which were difficult to measure by the experiments. (2)The boundary surface temperature between the cover gas layer and the rotating plug had the same tendencies and the calculated average temperatures on the boundary surface had good agreements with the experimental data. The average relative deviations from experimental values were less than 1.3%. (3)The natural convection of the cover gas enhanced the temperature distributions in the structure. The effects of thermal radiation on the heat transfer was relatively small and it can be neglected when the temperature of the heated aluminum disk is less than 400
C.
*
PNC TN9410 88-182, 70 Pages, 1988/12
The FLUSH, which has been developed for the purpose of calculating the interaction between thermal-hydraulics and thermal structural responses, was validated through the calculation of SPINTA and the comparizon of its experimental results. The results obtained are as follows; (1)The temperature distributions of the shield plug were able to simulate within 17% error of the experimental results, (2)The thermal stress distributions were normally calculated, (3)The iteration was converged in 0.27% of the norm of temperature. and it took about 122 minutes as the total CPU time of the calculation. These results show that even the large components of FBR can be analysed sufficiently and accurately in reasonable CPU time through the FLUSH.
*
PNC TN9410 88-181, 130 Pages, 1988/12
In order to evaluate the natural convection of Ar gas and structures interaction of the upper closure system of SNR300, the benchmark calculations were carried out at both Interatom GmbH and PNC in the same conditions. Three dimensional temperature distribution analysis of the cover gas and the shield plug, and thermal stress analysis of the shield plug were carried out in the different two cases of sodium temperature, 420C and 556C, using the Interface Program for Fluid and Structures Interaction, "FLUSH", at PNC. And following results were obtained; (1)The surface temperature on the upper plate was in the region of 106.6C
176.0C. These rather high values were caused by the ECC (Emergency Cooling Circuit) in the shield plug being performed as the heat source. (2)The maximum circumferential temperature difference was 81.0C, and the maximum thermal stress on the upper plate caused by that ununiform temperature difference was 37.6N/mm
. The technical information exchange meeting is going to be held in order to compare the PNC results with the SNR300 experimental data and the calculation results performed by using COMMIX/STINT code at Interaton.
; ; Naruse, Yuji
JAERI-M 8612, 30 Pages, 1979/12
no abstracts in English
; Ito, Hiroshi; Yoshida, Kenzo; Araki, Kunio
JAERI-M 6020, 33 Pages, 1975/03
no abstracts in English
*
Uraga Juko Giho, p.0 - 0, 1963/03
no abstracts in English
Nakayama, Masashi; Miura, Norihiko*; Ishida, Tomoko*; Takeda, Nobufumi*; Niunoya, Sumio*; Jo, Mayumi*
no journal, ,
no abstracts in English
Motoshima, Takayuki*; Usui, Tatsuya*; Sakamoto, Atsushi*; Niunoya, Sumio*; Ishida, Tomoko*; Miura, Norihiko*; Nakayama, Masashi; Ono, Hirokazu
no journal, ,
Japan Atomic Energy Agency is conducting the in-situ experiment for verification of performance of engineered barrier system in Horonobe Underground Research Laboratory. The purpose of this paper is reporting the mix design of high-workable low-alkaline plug concrete. This paper also reports the thermal stress analysis of plug concrete and results of in-situ thermal measurements.
Yoshikawa, Ryuji; Imai, Yasutomo*; Tanaka, Masaaki; Ohshima, Hiroyuki
no journal, ,
Accurate evaluation of three-dimensional temperature distribution is essentially required for the structural integrity analysis of large-sized straight tube steam generator (SG) of an advanced sodium-cooled fast reactor (SFR). Numerical simulation system TSG to analyze three-dimensional thermal-hydraulics in the straight tube SG has been developed. The three-dimensional simulation on sodium side by the CFD code FLUENT with porous body model was coupled with the multi-channel simulation on water/steam side. Fundamental validation of TSG code with the 1MWt straight tube SG test was performed. Temperature distribution of large-sized SG under the condition with plugged tubes was analyzed and the temperature deviation of heat transfer tubes was evaluated. Through the numerical simulation, local temperature increase near the plugged tubes was quantitatively evaluated, and the applicability of the simulation results to the structural integrity analysis of straight tube SG was indicated.
Tanaka, Tatsuya*; Toya, Naruhisa*; Okugi, Sakura*; Hashimoto, Shuya*; Matsui, Hiroya
no journal, ,
no abstracts in English
