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Fukui, Katsunori*; Hashiba, Kimihiro*; Sato, Toshinori; Kuwabara, Kazumichi; Takayama, Yusuke
JAEA-Research 2015-015, 61 Pages, 2015/11
JAEA-Research-2015-015.pdf:5.52MB
With respect to high-level radioactive waste disposal, knowledge of the long-term mechanical stability of shafts and galleries excavated in rock is required, not only during construction and operation but also over a period of thousands of years after closure. On the other hand, it is known that rock and the rock mass surrounding the disposal gallery shows time dependent behavior such as creep or the stress-relaxation. It becomes the issue in the stability evaluation of the disposal gallery to grasp the behavior. About this issue, we pushed forward research development. In the fiscal year of 2014, the creep test was continuously conducted and the total testing time exceeded 17 years. The testing equipment and procedure were examined to investigate the deformation, failure and time-dependency of rock under wet conditions and between room temperature and 100
C. The long-term strength of rock under triaxial stress state was researched with the aid of laboratory testing results and in situ stress measurement.
Fukui, Katsunori*; Hashiba, Kimihiro*; Sato, Toshinori; Sanada, Hiroyuki; Kuwabara, Kazumichi
JAEA-Research 2014-020, 50 Pages, 2014/11
JAEA-Research-2014-020.pdf:2.8MB
On the radioactive waste disposal, the long-term mechanical stability of shafts and galleries excavated in rock is required. Therefore, it is very important to understand the time-dependent behavior of rock mass for evaluating long-term mechanical stability. The purpose of this study is determining the mechanisms of time-dependent behavior of rock mass by precise testing, observation and measurement in order to develop methods for evaluating long-term mechanical stability of a rock mass. This report describes the results of the activities in fiscal year 2013. In Chapter 1, we described the overview and background of this study. In Chapter 2, the results of a long-term creep test on Tage tuff, started in fiscal year 1997 are described. In Chapter 3, the result of organization and analysis for time-dependent behavior of crystalline rock was described. In Chapter 4, for the drafting of in-situ test plan, examination of the numerical analysis technique of rock mass was carried out.
Aoyagi, Takayoshi*; *; Mihara, Morihiro; Okutsu, Kazuo*; Maeda, Munehiro*
JNC TN8400 2001-024, 103 Pages, 2001/06
JNC-TN8400-2001-024.pdf:8.84MB
In the disposal concept of TRU waste, concentrated disposal of wastes forms in large cross-section underground cavities is envisaged, because most of TRU waste is no-heat producing in spite of large generated volume as compared with HLW. In the design of engineered barrier system based on large cross-section cavities, it is necessary to consider the long-term mechanical process such as creep displacement of the host rock from the viewpoint of the stability of engineered barrier system. In this study, the long-term creep displacement of the host rock was calculated using the non-linear viscoelasticity model and the effects on the stability of engineered barrier system was evaluated. As a result, in the disposal concept of crystalline rock, no creep displacement occurred at the time after 1 milion year. On the other hand, in the disposal concept of sedimentary rock, creep displacement of 80
90mm occurred at the time after 1 milion year. Also, in this calculation, a maximum reduction of 45mm concerned with the thickness of buffer material was estimated. But these values resulted within allowance of design values. Therefore, these results show that the effects of the creep displacement on the stability of engieered barrier system would not be significant.
JNC TN8400 2001-008, 36 Pages, 2001/03
JNC-TN8400-2001-008.pdf:2.92MB
Research on geologic disposal of high-level radioactive waste(HLW) has been underway in many countries. Bentonite exhibiting a low permeability, high swelling property and high sorption capacity for many radioelements is proposed as a buffer material in many countlies. Recently, cementitious materials are considered as candidate matelials for the geologic disposal of high-level radioactive waste. As the pH and the Ca, Na, K contents of hyperalkaline pore water from the cementitious materials are high, this hyperalkaline pore water would alter the buffer material. The main aim of this study is to investigate the effect of alkaline pore water into the bentonite. Used materials are montmorillonite, albite and quartz composing bentonite. These minerals mixed in a constant ratio (1:1wt%) made to react to distilled water and the alkali solutions (pH11-13). These studies have been conducted at temperatures of 50 - 150C and run times of 10 - 200 day. XRD(X-ray powder diffraction) and SEM (Scanning Electron Microscopy) analyses were applied to studying the structure and quantitative data of each sample. From the result of this study, the main formed mineral of this experiment was analcime, which showed the tendency with a large amount of generation at a higher pH and temperature. Quantitative data of this study was conducted by X-ray powder diffraction method. THe order of the amount of the second analcime in each experiment is shown in the following. Montmorillonite and albite mixing test 
Montmorillonite test Montmorillonite and quartz mixing test Activation energies (E
) using the quantitative data of each test are shown in the following. (1)Montmorillonite test : 54.9kJ/mol (2)Montmorillonite and albite mixing test : 51.9kJ/mol (3)Montmorillonite and quartz mixing test : 59.6kJ/mol
; Makino, Hitoshi; Peter*
JNC TN8400 99-095, 69 Pages, 1999/12
MESHNOTE code was developed to evaluate the engineered barrier system in collaboration with QuantiSci. This code is used to simulate glass dissolution, diffusive transport of nuclides in the buffer material and release to surrounding host rock. MESHNOTE is a one-dimensional finite difference, code, which uses cylindrical co-ordinates for the solution of a radially symmetric diffusion problem. MESHNOTE has the followig characteristics: (1) MESHNOTE can solve for diffusive transport of nuclides through an annulus shaped buffer region while accounting for multiple decay chains, linear and non-linear sorption onto the buffer materials and elemental solubility limits; (2) MESHNOTE can solve for ingrowth of plural daughter nuclides from a singular parent nuclide (branching), and the ingrowth of a singular daughter nuclide from plural parent nuclides (rejoining); (3) MESHNOTE can treat the leaching of nuclide from the vitrified waste and the release of nuclide from buffer to surrounding rock, which are boundary conditions for migration in the buffer, basing on the phenomena; (4) MESHNOTE can treat principal parameters (e.g. solubility and distribution coefficient) relevant to nuclide migration as time and space-dependence parameters; (5) The time stepping scheme in MESHNOTE is controlled by tolerance defined by the user. The time stepping will increase automatically while checking the accuracy of the numerical solution. The conceptual model, the mathematical model and the numerical implementation of the MESHNOTE code are described in this report and the characteristic functions of MESHNOTE are verified by comparing with analytical solutions or simulations produced with other calculation codes.
; Kubota, Mitsuru*; *; *
JNC TN8430 98-001, 12 Pages, 1998/11
Leaching experiments were performed with simulated nuclear waste glasses in silica saturated water at 90C for periods of 28 days in order to clarify the alteration mechanism of waste glass under silica saturation. After leaching experiments, the depth profiles of some elements of glass surfaces were measured by ERDA and ESCA. Depth profiles of H and soluble elements, such as B and Na, for the simulated nuclear waste glass after leaching experiments demonstrated the formation of glass hydration layer and the depletion of soluble elements in the glass hydration layer. The results suggest that alteration of nuclear waste glass under silica saturation is attributed to glass hydration and the leaching behaviour of soluble elements depends on the growth of the glass hydration layer.
; Kano, Shigeki; ; Tachi, Yoshiaki; Hirakawa, Yasushi; Yoshida, Eiichi
PNC TN9410 98-021, 68 Pages, 1998/02
Engineering ceramics have excellent properties such as high strength, high hardness and high heat resistance compared with metallic matelials. To apply the ceramic in fast reactor environment, it is necessary to evaluate the sodium compatibility and the influence of sodium on the mechanical properties of ceramics. In this study, the influence of high temperature sodium on the mechanical properties of sintered ceramics of conventional and high purity Al
O
, SiC, SiAlON, AlN and unidirectional solidified ceramics of AlO/YAG eutectic composite were investigated by means of flexure tests. Test specimens were exposed in liquid sodium at 823K and 923K for 3.6Ms. There were no changes in the flexural strength of the conventional and high purity AlO, AlN and AlO/YAG eutectic composite after the sodium exposure at 823K. On the contrary, the decrease in the flexural strength was observed in SiC and SiAlON. After the sodium exposure at 923K, there were also no changes in the flexural strength of AlN and AlO/YAG eutectic composite. In the conventional and high purity AlO and SiC, the flexural strength decreased and signs of grain boundary corrosion were detected by surface observation. The flexural strength of SiAlON after the sodium exposure at 923K increased instead of severe corrosion. In the specimens those showed no changes in the flexural strength, further exposure in sodium is needed to verify whether the mechanical properties degrade or not. For SiAlON, it is necessary to clarify the reason for the increased strength after the sodium exposure at 923K.
; Katsuragi, Satoru; Tone, Tatsuzo
JAERI-M 4503, 99 Pages, 1971/07
no abstracts in English
Fujiwara, Kenso; Sasaki, Takayuki*; Matoba, Daisuke*; Iijima, Kazuki
no journal, ,
Depth profiles of radionuclides atmospheric discharged to the environment of Fukushima was obtained nearby the Fukushima Daiichi Nuclear Power Station (FDNPS). According to the profiles, it was suggested that deposition of Pu-238 and Cs-137 was mostly caused by atmospheric discharge by the FDNPS accident, however, in the case of Pu-239+240, it is necessary to consider the effect of global fallout. For understanding on the migration behaviour of radionuclide, it is necessary to obtain the profiles attributed by the accident and global fallout. In this study, in order to clarify the profile of soil, time dependence of depth profile of Cs and Pu in the soil outside of FDNPS site was measured. Based on the results of this study, it was confirmed that Pu and Cs will not be transferred significantly in the soil over several years. In the case of Pu, concentration of fallout and discharge was calculated. This data is considered to be useful for the safety assessment of waste in FDNPS, especially, measurement of other nuclides can also show the utility of transport ratio based on Cs concentration.
Tamaki, Hitoshi
no journal, ,
no abstracts in English