Benbow, S. J.*; Kawama, Daisuke*; Takase, Hiroyasu*; Shimizu, Hiroyuki*; Oda, Chie; Hirano, Fumio; Takayama, Yusuke; Mihara, Morihiro; Honda, Akira
Crystals (Internet), 10(9), p.767_1 - 767_33, 2020/09
Details are presented of the development of a coupled modeling simulator for assessing the evolution in the near-field of a geological repository for radioactive waste disposal where concrete is used as a backfill. The simulator uses OpenMI, a standard for exchanging data between simulation software programs at run-time, to form a coupled chemical-mechanical-hydrogeological model of the system. The approach combines a tunnel scale stress analysis finite element model, a discrete element model for accurately modeling the patterns of emerging cracks in the concrete, and a finite element and finite volume model of the chemical processes and alteration in the porous matrix and cracks in the concrete, to produce a fully coupled model of the system. Combining existing detailed simulation software in this way with OpenMI has the benefit of not relying on simplifications that might be necessary to combine all of the modeled processes in a single piece of software.
Goto, Akira; Murakami, Masaki*; Sakai, Ryutaro*; Terusawa, Shuji*; Sueoka, Shigeru
JAEA-Review 2020-003, 60 Pages, 2020/03
One of the natural phenomena that may affect the geological disposal system are earthquake and fault activity. Fault displacement due to the earthquake and fault activity will be considered the direct effects. In addition to it, it is necessary to consider the secondary effects include secondary faults formed by the seismic fault activity as well as spring water and mud volcanoes that are generated by fluid movement attributed to the fault activity. This paper introduces previous studies performed focused on the hydraulic effects (spring water and mud volcanoes) and mechanical effects (secondary faults) in order to understand the effects of these secondary phenomena on the geological disposal system. We were able to collect 142 literatures from Japan and overseas by searching for related keywords in Japanese and English. As a result, we compiled case studies of each secondary impact. From the viewpoint of geological disposal, we extracted the following issues for future research and development. As for the sump water induced by earthquakes and faulting, accumulation of information related to its mechanism, affected area, and activity history is required. As for the mud volcanoes, reviewing of the mechanism of anomalous pore water pressure that causing the formation, also development of estimation technique are required. And for the secondary faults, accumulation of the detailed spatial distribution and reviewing of formation mechanism are required.
Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro
Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00360_1 - 19-00360_13, 2020/03
It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (BC) as a control rod material and stainless steel (SS) during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan because the BC-SS eutectic relocation behavior has a large uncertainty in the reactivity history based on a simple calculation. A physical model simulating the eutectic melting reaction and relocation was developed and implemented into a severe accident simulation code. The developed model must be validated by using test data. To validate the physical model, therefore, the visualization tests of SS-BC eutectic melting reaction was carried out by contacting SS melts of several kg with a BC pellet heated up to about 1500 C. The tests have shown the eutectic reaction visualization as well as freezing and relocation of the BC-SS eutectic in upper part of the solidified test piece due to the density separation. Post-test material analyses by using X-ray diffraction and transmission electron microscope techniques have indicated that FeB appeared at the BC-SS contact interface and (Fe,Cr)B at the top surface of the test piece. Glow discharge optical emission spectrometry has been applied to quantitative analysis of boron concentration distributions. The boron concentration was high at the upper surface and near the original position of the BC pellet.
Kajita, Yuya*; Fukuda, Shoma*; Sueoka, Shigeru; Hasebe, Noriko*; Tamura, Akihiro*; Morishita, Tomoaki*; Tagami, Takahiro*
Fisshion, Torakku Nyusureta, (32), p.6 - 7, 2019/12
no abstracts in English
Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa
Nuclear Technology, 205(9), p.1154 - 1163, 2019/09
An electromagnetic-levitation technique performed in a static magnetic field was used to measure the density, surface tension, normal spectral emissivity, heat capacity, and thermal conductivity of molten 316L stainless steel (SS316L) and SS316L that contained 5mass%BC. The addition of 5mass%BC to SS316L yielded reductions of 111 K, 6%, 19%, and 6% in the liquidus temperature, density, normal spectral emissivity, and thermal conductivity at the liquidus temperature of SS316L, respectively. The heat capacity increased by 5% with this addition. Although the 5mass%BC addition had no clear effect on the surface tension, sulfur dissolved in the SS316L resulted in a significant decrease in the surface tension.
Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro; Kikuchi, Shin; Emura, Yuki; Kamiyama, Kenji; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; et al.
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.418 - 427, 2019/09
Eutectic reactions between boron carbide (BC) and stainless steel (SS) as well as its relocation are one of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors. Since such behaviors have never been simulated in CDA numerical analyses, it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study is focusing on BC-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in the range from solid to liquid state. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies by 2017. Specific results in this paper is boron concentration distributions of solidified BC-SS eutectic sample in the eutectic melting experiments, which would be used for the validation of the eutectic physical model implemented into the computer code.
Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Nozawa, Takashi*; Sato, Daiki*; Murakami, Nozomu*; et al.
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.206 - 216, 2019/09
After the nuclear accident at Fukushima Daiichi Power Plant, research and development (R&D) program for establishing technical basis of accident-tolerant fuel (ATF) started from 2015 in Japan. Since then, both experimental and analytical studies necessary for designing a new light water reactor (LWR) core with ATF candidate materials are being conducted within the Japanese ATF R&D Consortium for implementing ATF to the existing LWRs, accompanying with various technological developments required. Until now, we have accumulated experimental data of the candidate materials by out-of-pile tests, developed fuel evaluation codes to apply to the ATF candidate materials, and evaluated fuel behavior simulating operational and accidental conditions by the developed codes. In this paper, the R&D progresses of the ATF candidate materials considered in Japan are reviewed based on the information available such as proceedings of international conference and academic papers, providing an overview of ATF program in Japan.
Onoe, Hironori; Kosaka, Hiroshi*; Matsuoka, Toshiyuki; Komatsu, Tetsuya; Takeuchi, Ryuji; Iwatsuki, Teruki; Yasue, Kenichi
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 26(1), p.3 - 14, 2019/06
In this study, it is focused on topographic changes due to uplift and denudation, also climate perturbations, a method which is able to assess the long-term variability of groundwater flow conditions using the coefficient variation based on some steady-state groundwater flow simulation results was developed. Spatial distribution of long residence time area which is not much influenced due to long-term topographic change and recharge rate change during the past one million years was able to estimate through the case study of the Tono area, Central Japan. By applying this evaluation method, it is possible to identify the local area that has low variability of groundwater flow conditions due to topographic changes and climate perturbations from the regional area quantitatively and spatially.
Kaji, Yoshiyuki; Nemoto, Yoshiyuki; Nagatake, Taku; Yoshida, Hiroyuki; Tojo, Masayuki*; Goto, Daisuke*; Nishimura, Satoshi*; Suzuki, Hiroaki*; Yamato, Masaaki*; Watanabe, Satoshi*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
In this research program, cladding oxidation model in SFP accident condition, and numerical simulation method to evaluate capability of spray cooling system which was deployed for spent fuel cooling during SFP accident, have been developed. These were introduced into the severe accident codes such as MAAP and SAMPSON, and SFP accident analyses were conducted. Analyses using Computational Fluid Dynamics (CFD) code were conducted as well for the comparison with SA code analyses and investigation of detail in the SFP accident. In addition, three-dimensional criticality analysis method was developed as well, and safer loading pattern of spent fuels in pool was investigated.
Machida, Hideo*; Arakawa, Manabu*; Wakai, Takashi
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
This paper describes the effect of local plastic component on J-integral and crack opening displacement (COD) evaluation of a circumferential penetrated crack, applicable to the leak before break (LBB) assessment for sodium cooled fast reactor (SFR) components. J-integral COD evaluation methods are generally formulated as a summation of elastic and plastic components, and so far many evaluation formulae based on these two components have been proposed. However, strictly, the plastic component consists of local plastic and fully plastic components. Many of the conventional evaluation methods often consider only the fully plastic component as the plastic component. The reason for this is that the effect of the local plastic component is much smaller than that of the fully plastic component excluding materials with extremely small work hardening. In contrast, for materials with high yield stress and small work hardening, such as modified 9Cr-1Mo steel which is one of the candidate materials for SFR piping, the effect of the local plastic component on J-integral and COD cannot be ignored. Therefore, the authors propose formulae taking the effect of local plastic component on J-integral and COD into account, based on finite element analysis (FEA) results, so that it is easy to apply to crack evaluation. The formulae will be employed in the guidelines on LBB assessment for SFR components published from Japan Society of Mechanical Engineers (JSME).
Shirasu, Noriko; Suzuki, Akihiro*; Nagae, Yuji; Kurata, Masaki
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05
High temperature interaction tests between UO and Zr were performed at around 2173 K, to make clear the UO/ -Zr(O) interaction and the mechanism of degradation, for developing the improved models for advanced severe accident analysis codes. A Zr plate was inserted in a UO crucible, and heat treated at 2173 K in stream of Ar. After the heat-treatment, the samples were subjected to surface microanalysis. The middle region of Zr sample shows streak-like structures which are extended towered the top. It is confirmed that the streak-like structures were mainly consist of U from the EDX results, and the structures revealed that the U-rich phase was liquid during the heat-treatment. It seems that the U-rich liquid grew selectively toward the area where the oxygen concentration was low.
Kofuji, Hirohide; Watanabe, So; Takeuchi, Masayuki; Suzuki, Hideya; Matsumura, Tatsuro; Shiwaku, Hideaki; Yaita, Tsuyoshi
Progress in Nuclear Science and Technology (Internet), 5, p.61 - 65, 2018/11
Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*
Nippon Kikai Gakkai 2018-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2018/09
According to the fitness for service code of Sodium-Cooled fast Reactor (SFR), the volumetric tests as in-service inspection can be replaced with continuous leak monitoring, where the Leak Before Break (LBB) is demonstrated, because the primary stress caused by internal pressure is not significant in SFR components. Basically, if the detectable crack length and the penetrated crack length are sufficiently smaller than the unstable critical crack length, it can be concluded that LBB is successfully demonstrated. The authors had already proposed a simplified method to calculate the penetrated crack length both of the circumferential and axial cracks in the pipe as a function of pipe geometry, fatigue crack growth characteristics and loading conditions. However, some problems in the method have been pointed out in the process of the reviewing by the JSME code committee. This study describes an improved method to calculate the penetrated crack length.
Kokubo, Hiroki*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Yamano, Hidemasa
Nippon Kinzoku Gakkai-Shi, 82(10), p.400 - 402, 2018/09
It is important to obtain the viscosity of a mixed alloy consisting of molten stainless steel and boron-carbide (SUS316L + BC alloy) for the improvement of severe accident assessment methodology for sodium-cooled fast reactors. In this study, the viscosities of the molten nickel (Ni) and stainless steel (SUS316L) were measured by the oscillating crucible method to confirm the performance of the viscosity measurement apparatus as a first step. The viscosities of molten Ni and SUS316L melts were measured up to 1823 K. It was found that the measured viscosity values of molten Ni and SUS316L were estimated from the deviation of the experimental data, were 4% and 3%, respectively. It was also found that those of molten Ni and SUS316L were close to those of the literature values of molten Ni and similar composite stainless steels. Moreover, we tentatively measured the viscosity of molten SUS316L-5 mass%BC alloy. The fitted results of the viscosity for molten Ni and SUS316L were obtained.
Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Yanagihara, Seiji*; Suzuki, Ryosuke*; Matsubara, Masaaki*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07
This paper studies crack opening displacement (COD) evaluation methods used in Leak-Before-Break (LBB) assessment of Sodium cooled Fast Reactor (SFR) pipe. For SFR pipe, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. The sodium pipes are made of ASME Gr.91 (modified 9Cr-1Mo steel). Thickness of the pipes is small, because the internal pressure is very small. Modified 9Cr-1Mo steel has a relatively large yield stress and small work hardening coefficient comparing to the austenitic stainless steels which are currently used in the conventional plants. In order to assess the LBB behavior of the sodium pipes made of modified 9Cr-1Mo steel, the coolant leak rate from a through wall crack must be estimated properly. Since the leak rate is strongly related to the crack opening displacement (COD), an appropriate COD assessment method must be established to perform LBB assessment. However, COD assessment method applicable for SFR pipes - having thin wall thickness and made of small work hardening material - has not been proposed yet. Thus, a COD assessment method applicable to such a pipe was proposed in this study. In this method, COD was calculated by classifying the components of COD; elastic, local plastic and fully plastic. In addition, the verification of this method was performed by comparing with the results of a series of four-point bending tests using modified 9Cr-1Mo steel pipe having a circumferential through wall notch. As a result, in some cases, COD were over-estimated especially for large cracks. Although the elastic component of COD is still over-estimated for large cracks, leak evaluation from small cracks is much more important in LBB assessment. Therefore, this study recommends that only the elastic component of COD should be adopted in LBB assessment of SFR pipes.
Niwa, Masakazu; Nomura, Katsuhiro; Hiura, Yuki
JAEA-Review 2018-010, 40 Pages, 2018/04
In the Japanese Islands, coastal area can be proposed as an investigation site for geological disposal of high-level radioactive wastes. For an assessment of fault activity in coastal area, offshore surveys such as acoustic profiling and boring should be examined as well as inland surveys. In addition, adequate understanding spatial distributions and characteristics of faults in the coastal area of Japan will contribute to safety assessment for the geological disposal in such area. In this report, we collected and compiled previous studies focused on spatial distribution, continuity, timing of displacement and recurrence interval of faults near coastline, specifically faults along or across a boundary between land and sea, and technologies of survey and assessment for them.
Nakano, Hiroko; Hirota, Noriaki; Shibata, Hiroshi; Takeuchi, Tomoaki; Tsuchiya, Kunihiko
Mechanical Engineering Journal (Internet), 5(2), p.17-00594_1 - 17-00594_12, 2018/04
no abstracts in English
Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1014 - 1019, 2018/04
In this study, densities, surface tensions, normal spectral emissivities, heat capacities and thermal conductivities of molten SUS316L and SUS316L containing 5mass%-BC were measured by the electromagnetic levitation technique in a static magnetic field.
Hourcade, E.*; Mihara, Takatsugu; Dauphin, A.*; Dirat, J.-F.*; Ide, Akihiro*
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.556 - 561, 2018/04
In the framework of the French-Japanese agreement signed in 2014, CEA, AREVA NP, JAEA, and MHI/MFBR is jointly performing components design of ASTRID such as Decay Heat Removal Systems (DHRS). This paper is giving an update concerning ASTRID DHR strategy with description of reference architecture evolution and project objectives. In particular, new developments were made for DHR during normal shutdown and role of Ex-Vessel system. A special focus is made on design process of automatic shutter to hydraulically connect Hot Plenum and cold plenum to enhance primary vessel natural convection.
Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Kikuchi, Koichi*
Proceedings of ASME Symposium on Elevated Temperature Applications of Materials for Fossil, Nuclear, and Petrochemical Industries, 7 Pages, 2018/04
A simplified J-integral evaluation method applicable to unstable failure analysis in Leak Before Break (LBB) assessment of Sodium-cooled Fast Reactor (SFR) in Japan was proposed. Mod.9Cr-1Mo steel is supposed to be a candidate material for the coolant systems of SFR in Japan. This steel has relatively high yield strength and poor fracture toughness comparing to those of conventional austenitic stainless steels. In addition, SFR pipe has small thickness and large diameter. As a J-integral evaluation method for circumferential through-wall crack in a cylinder, EPRI has proposed a fully plastic solution method. However, the geometry of SFR pipe and material characteristics of Mod.9Cr-1Mo steel exceed the applicable range of EPRI's method. Therefore, a series of elastic, elasto-plastic and plastic finite element analyses (FEA) were performed for a pipe with a circumferential through-wall crack to propose a J-integral evaluation method applicable to such loading conditions. J-integrals obtained from the FEA were resolved into elastic, local plastic and fully plastic components. Each component was expressed as a function of analytical parameter, such as pipe geometries, crack size, material characteristics and so on. As a result, a simplified J-integral evaluation method was proposed. The method enables to conduct 2 parameter failure analysis using J-integral without any fracture mechanics knowledge.