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Onoe, Hironori; Ishibashi, Masayuki*; Ozaki, Yusuke; Iwatsuki, Teruki
International Journal of Rock Mechanics and Mining Sciences, 144, p.104737_1 - 104737_14, 2021/08
Times Cited Count：4 Percentile：43.25(Engineering, Geological)In this study, we investigated the methodology of modeling for fractured granite around the drift at a depth of 500 m in the Mizunami Underground Laboratory, Japan as a case study. As a result, we developed the fracture modeling method to estimate not only geological parameters of fractures but also hydraulic parameters based on the reproducibility of trace length distribution of fractures. By applying this modeling method, it was possible to construct a Discrete Fracture Network (DFN) model that can accurately reproduce the statistical characteristics of fractures.
Uwaba, Tomoyuki; Yokoyama, Keisuke; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*; Pelletier, M.*
Nuclear Engineering and Design, 359, p.110448_1 - 110448_7, 2020/04
Times Cited Count：1 Percentile：11.8(Nuclear Science & Technology)Coupled computer code analyses of irradiation performance of axially heterogeneous mixed oxide (MOX) fuel elements with high burnup in a fast reactor were conducted. Post-irradiation experiments revealed local concentration of Cs near the interfaces between MOX fuel and blanket columns including the internal blanket of the fuel elements as well as an increase in their cladding diameters. The analyses indicated that the local Cs concentration occurred as a result of Cs axial migration from the MOX fuels toward the blanket pellets near the interfaces. Swelling of the blanket pellets induced by the formation of low-density Cs-U-O compound was not sufficient to cause pellet-to-cladding mechanical interaction (PCMI). The PCMI analyzed in the MOX fuel column regions was insignificant, and the cladding diameter increases were caused mainly by void swelling in cladding and irradiation creep due to fission gas pressure.
Ohgama, Kazuya; Oki, Shigeo; Kitada, Takanori*; Takeda, Toshikazu*
Proceedings of 21st Pacific Basin Nuclear Conference (PBNC 2018) (USB Flash Drive), p.942 - 947, 2018/09
Onoe, Hironori; Yamamoto, Shinya*; Kohashi, Akio; Ozaki, Yusuke; Sakurai, Hideyuki*; Masumoto, Kiyoshi*
JAEA-Research 2018-003, 84 Pages, 2018/06
JAEA-Research-2018-003.pdf:17.44MB
In this study, numerical experiments considered hydrogeological structures, which has high heterogeneity around the Mizunami Underground Research Laboratory and inverse analysis using in-situ data were carried out. The results showed that concentration of hydrogeological structure to be estimated and location of monitoring point is important for application of inverse analysis. Furthermore, it is concluded that inverse analysis using hydraulic response due to pumping test is effective for hydrogeological characterization.
Aoyagi, Kazuhei; Kamemura, Katsumi*; Nago, Makito*; Sugawara, Kentaro*; Matsubara, Makoto*
Proceedings of ITA-AITES World Tunnel Congress 2017 (WTC 2017) (USB Flash Drive), 10 Pages, 2017/06
An in situ stress state is one of the important factors in the design of deep underground facility of high-level radioactive waste disposal repository. This study establishes a practical and effective method for estimating in situ stress state on the basis of the measured convergence during gallery excavation. The convergence was measured in various directions of the loop gallery at 350m depth of the Horonobe Underground Research Laboratory; this allows determination of the stress state corresponding to the rock mass deformation behavior in an approximately 120 m* 200 m area. To estimate in situ stress state around that area, a back analysis method considering the existence of faults and fractures around the gallery was developed. The analyzed results showed a good agreement with the trend of in situ stress state estimated from hydraulic fracturing method.
Onoe, Hironori; Saegusa, Hiromitsu; Takeuchi, Ryuji
Doboku Gakkai Rombunshu, C (Chiken Kogaku) (Internet), 72(1), p.13 - 26, 2016/01
The Japan Atomic Energy Agency is conducting the Mizunami Underground Research Laboratory (URL) project in Mizunami, Gifu, in order to establish scientific and technical basis for geological disposal of high-level radioactive waste. This paper comprehensively describes the result of groundwater flow modeling using data of hydraulic responses and hydrochemical changes due to URL construction. Technical know-how and methodology of hydrogeological monitoring and groundwater flow modeling were presented for characterization of hydraulic heterogeneities in fractured crystalline rock. Furthermore, effectivity of data acquisition of hydrochemical changes in groundwater for validation of result of groundwater flow modeling was indicated.
Takahashi, Satoshi*; Okuno, Hiroshi; Miyoshi, Yoshinori
JAERI-Tech 2005-056, 51 Pages, 2005/09
JAERI-Tech-2005-056.pdf:2.92MB
In the heterogeneous system of the mixed oxide fuel of uranium and plutonium, hereafter, MOX fuel, it was investigated whether the system could be modeled as a homogeneous system on the conditions which dealt with the MOX fuel of particle diameter 0.02mm or less in MOX fuel fabrication facilities in Japan. The infinite multiplication factor of the homogeneous system of the MOX fuel was first calculated, and the optimum moderation condition over the each ratio of PuO
was determined. It was verified that carried out critical calculation for the heterogeneous system of the MOX fuel in which the spherical fuel diameter in a cube unit cell increased, and an atomic number ratio of hydrogen to heavy metal fixed conditions, and the probability for neutrons to escape resonance by a spherical fuel diameter no less than 0.1mm, and analyzed critical conditions etc. using a contiguous energy Monte Carlo code MVPII and JENDL3.3. The details of these calculations are reported. These results are expected to be quoted in a revised edition of "Nuclear Criticality Safety Handbook."
Hidaka, Akihide*; Kudo, Tamotsu; Kida, Mitsuko; Fuketa, Toyoshi
JAERI-Research 2005-001, 67 Pages, 2005/02
JAERI-Research-2005-001.pdf:3.38MB
In the VEGA program to investigate radionuclides release from irradiated fuel during severe accidents, the analyses are being performed with VICTORIA2.0 code for comprehensive understanding of radionuclides release and transport phenomena. The VEGA-1 and -3 tests were analyzed in the present study. The correlation for Cs diffusion coefficient in fuel grain obtained from VEGA-1 was applied to the release analysis of VEGA-3. The calculated release of Cs agreed well with the measurement. The correlation was applied to subsequent Cs transport and deposition analyses. The calculation underpredicted the total mass of Cs deposited onto the test apparatuses because nucleation of aerosol and its growth were underestimated due to the consideration of aerosol nucleation originated only from released FP in VICTORIA2.0. A sensitivity analysis with aerosol seeds for heterogeneous nucleation showed a reasonable agreement with the measured Cs distribution. It turned out that additional aerosol seeds besides the released FP be considered when the VICTORIA2.0 code is applied to the VEGA test analyses.
Arai, Yasuo; Pillon, S.*
Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles (CD-ROM), 9 Pages, 2004/06
no abstracts in English
Sono, Hiroki; Fukaya, Yuji; Yanagisawa, Hiroshi; Miyoshi, Yoshinori
JAERI-Tech 2003-065, 61 Pages, 2003/07
JAERI-Tech-2003-065.pdf:3.11MB
A series of critical experiments using a heterogeneous core of the Static Experiment Critical Facility (STACY) in the Japan Atomic Energy Research Institute is planned in F.Y. 2003. In the experiment, the core is composed of uranyl nitrate solution (
U enrichment 6 wt%) and 333 pins of uranium dioxide (U enrichment 5 wt%) loaded in lattice-pitch of 1.5 cm. Prior to the experiment, neutronic characteristics are analyzed to evaluate neutronic safety and criticality limitations of the core. The analyzed items are the parameters on criticality, reactivity and reactor shutdown margins. In the analyses, a Monte Carlo code, MVP, and a neutronics code system, SRAC, have been used with an evaluated nuclear data library, JENDL-3.3. By using the calculated characteristics, simplified equations to interpolate these values and criticality limitations of the core are evaluated. It has been also confirmed that the reactor shutdown margins will comply with safety criteria under all fuel conditions in the experiments.
Yamamoto, Toshihiro; Miyoshi, Yoshinori; Kikuchi, Tsukasa*; Watanabe, Shoichi
Journal of Nuclear Science and Technology, 39(7), p.789 - 799, 2002/07
Times Cited Count：5 Percentile：34.51(Nuclear Science & Technology)no abstracts in English
Tanaka, Satoru*
JAERI-Tech 2002-016, 110 Pages, 2002/03
JAERI-Tech-2002-016.pdf:5.52MB
no abstracts in English
*;
JNC TN8400 2001-027, 131 Pages, 2001/11
In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostastical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostastical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostastical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report.
Chinju, H.*; Nagasaki, Shinya*; Tanaka, Satoru*; Sakamoto, Yoshiaki; Takebe, Shinichi; Ogawa, Hiromichi
Journal of Nuclear Science and Technology, 38(8), p.645 - 654, 2001/08
Times Cited Count：3 Percentile：27.07(Nuclear Science & Technology)no abstracts in English
*; Kanazawa, Yasuo*;
JNC TN8400 2001-012, 69 Pages, 2001/04
JNC-TN8400-2001-012.pdf:6.87MB
On understanding the radionuclide transport in natural barrier in radioactive waste isolation research, the macroscopic dispersion in heterogeneous permeability field in the underground rock is regarded as an important process. Therefore, we have conducted lots of tracer experiments by the MACRO II facility with an artificially constructed heterogeneous permeability field. In order to study the scale dependence of dispersion coefficients in case of laboratory experiments, we placed the flow cell horizontally, and conducted injection-withdraw tracer experiment with a single well. We have conducted I5 cases experiments. These cases were prepared by changing a position of single well and the injection-withdraw time. At each position we have conducted 9 cases and 6 cases experiments. In this report, we evaluated the macroscopic dispersion coefficients by the fitting of analytical solution to breakthrough curve measured by the 15 cases pumping tracer experiment. Consequently, we could evaluate the dispersion coefficients for 12 cases of 15 cases. Then, we discussed the relation between a injection-withdraw flow rate and a property of heterogeneous media and dispersion coefficient. The conclusions obtained from the results of the evaluation are summarized as follows, (1)It was found that the macroscopic dispersion coefficients tend to be increased with increase of the average radius of tracer front spread around a single well. (2)We have conducted any experiments with s single well settled at two positions. In case of that there is low permeability around a single well, we found dispersion coefficients are large. In case of that there is high permeability around a single well, we found dispersion coefficients are small. (3)In three cases that we could not evaluate because of incorrect accuracy of fitting, we have found it possible that there is some points that dispersion coefficients were strikingly small in tracer front.
*; *; *; *
JNC TJ8400 2000-006, 232 Pages, 2000/05
JNC-TJ8400-2000-006.pdf:7.75MB
In this research, simulations with some parameters which characterize ground water flow and the reliability evaluation for the expansion of the calculation method of groundwater flow were carried out by using the radionuclide transport computations in nearfield heterogeneous porous media. Concretely contents are follows: (1)With the series of calculation method for three-dimensional saturated/unsaturated groundwater flow and one-dimensional radionuclide transport. the computational analyses with the parameters used in JNC report in 2000 was carried out and the influence of the different input flux was evaluated. (2)The examination of the application for the different ways of inverse laplace transformation which is used in one-dimensional radionuclide tansport analysis code "MATRICS" was carried out. (3)The examination of the application of multi-element "MATRICS" (m-MATRICS) for radionuclide transport computations in nearfield heterogeneous porous media was carried out. (4)The series of calculation methods from three-dimensional saturated/unsaturated ground water flow simulation code to one-dimensional radionuclide transport simulation code was integrated.
; Sato, Wakaei*;
JNC TN9400 2000-037, 87 Pages, 2000/03
JNC-TN9400-2000-037.pdf:3.48MB
ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and 
-value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...
*
JNC TJ9400 2000-005, 182 Pages, 2000/03
JNC-TJ9400-2000-005.pdf:4.74MB
The SLAROM code, performing fast reactor cell calculation based on a deterministic methodology, has been revised by adding the universal module PEACO of generating Ultra-fine group neutron spectra. The revised SLAROM, then, was utilized for evaluating reaction rate distributions in ZPPR-13A simulated by a 2-dim RZ homogeneous model, although actually ZPPR-13A composed of radial heterogereous cells. The reaction rate distributions of ZPPR-13A were also calculated by the code MVP, that is a continuous energy Monte Carlo calculation code based on a probabilistic methodology. By coparing both results, it was concluded that the module PEACO has excellent capability for evaluating highly accurate effective cross sections. Also it was proved that the use of a new fine group cross section library set (next generation set), reflecting behavior of cross sections of structural materials, such as Fe and O, in the fast neutron energy region, is indispensable for attaining a better agreement within 1% between both calculation methods. Also, for production of a next generation set of group cross sections, the code NJOY97.V107 was added to the group cross section production system and both front and end processing parts were prepared. This system was utilized to produce the new 70 group JFS-3 library using the evaluated nuclear data library JENDL-3.2. Furthermore, to confirm the capability of this new group cross section production system, the above new JFS-3 library was applied to core performance analysis of ZPPR-9 core with a 2-dim RZ homogeneous model and analysis of heterogeneous cells of ZPPR-9 core by using the deterministic method. Also the analysis using the code MVP was performed. Bycoaparison of both results the following conclusion has been derived; the deterministic method, with the PEACO module for resonance cross sections, contributes to improve accuracy of predicting reaction rate distributions and Na void reactivity in fast reactor cores. And it ...
Sawada, Atsushi; Ijiri, Yuji; *; Watari, Shingo
JNC TN8400 99-093, 58 Pages, 1999/11
This paper decribes the results of PA studies considering heterogeneous fracture characteristics, for the purpose of contributing for the performance assessment of the natural barrier system PA in H12 report (The second progress report on research and development for the Geological Disposal of HLW in Japan). In this study, 3-D discrete fracture network mode1 (DFN) and 1-D multiple pathways model is applied for 100m scale of rock block. Although nuclide release rate calculated by DFN are widely distributed among the realizations, it is shown that several tens realizations are enough number to understand the stochastic characteristics of the nuclide release. From the data uncertainty analysis, there are no significant effects for the nuclide retardation in fracture geometry parameters such as fracture radius, density and etc. 1-D multiple pathways model is developed with focusing on the heterogeneity of the transmissivity, which has a large effect to the nuclide retardation effects. The nuclide release rate calculated by using 1-D multiple pathways model approximates to the results of DFN. This result also shows that the relatively large fractures/faults that connects disposal tunnel and downstream faults have an important role for performance assessment in natural barrier system.
Oigawa, Hiroyuki
JAERI-Research 98-061, 22 Pages, 1998/11
JAERI-Research-98-061.pdf:0.76MB
no abstracts in English
