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Uwaba, Tomoyuki; Ito, Masahiro*; Ishitani, Ikuo*
JAEA-Technology 2023-006, 36 Pages, 2023/05
JAEA-Technology-2023-006.pdf:3.45MB
The BAMBOO code developed by the Japan Atomic Energy Agency is a computer code to analyze fuel pin bundle deformation in a fast reactor wire-spaced type fuel pin bundle subassembly. In this study we developed an analysis model to consider friction at the contact points between adjacent fuel pins, and at these between outermost fuel pins and a duct that are due to bundle-duct interaction. This model deals with friction forces at contact points in the contact and separation analysis of the code, and employs a convergent calculation where contact forces are gradually determined to avoid numerical instability when the friction occurs. Analyses of BAMBOO with the model showed very slight effects on the onset of contact between outer most pins and a duct, and on directions of pin displacements, within the range of practical friction coefficients.
Yamamoto, Kazami; Kinsho, Michikazu; Hayashi, Naoki; Saha, P. K.; Tamura, Fumihiko; Yamamoto, Masanobu; Tani, Norio; Takayanagi, Tomohiro; Kamiya, Junichiro; Shobuda, Yoshihiro; et al.
Journal of Nuclear Science and Technology, 59(9), p.1174 - 1205, 2022/09
Times Cited Count：6 Percentile：84.97(Nuclear Science & Technology)In the Japan Proton Accelerator Research Complex, the purpose of the 3 GeV rapid cycling synchrotron (RCS) is to accelerate a 1 MW, high-intensity proton beam. To achieve beam operation at a repetition rate of 25 Hz at high intensities, the RCS was elaborately designed. After starting the RCS operation, we carefully verified the validity of its design and made certain improvements to establish a reliable operation at higher power as possible. Consequently, we demonstrated beam operation at a high power, namely, 1 MW. We then summarized the design, actual performance, and improvements of the RCS to achieve a 1 MW beam.
Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi
Nuclear Technology, 207(8), p.1280 - 1289, 2021/08
Times Cited Count：3 Percentile：35.51(Nuclear Science & Technology)Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.
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：12.16(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.
Uwaba, Tomoyuki; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*
Nuclear Engineering and Design, 331, p.186 - 193, 2018/05
Times Cited Count：4 Percentile：38.58(Nuclear Science & Technology)A computer code for the analysis of the overall irradiation performance of a fast reactor mixed-oxide (MOX) fuel element was coupled with a specialized code for the analysis of fission product cesium behaviors in a MOX fuel element. The coupled code system allowed for the analysis of the radial and axial Cs migrations, the generation of Cs chemical compounds and fuel swelling due to Cs-fuel-reactions in association with the thermal and mechanical behaviors of the fuel element. The coupled code analysis was applied to the irradiation performance of a fast reactor MOX fuel element attaining high burnup for discussion on the axial distribution of Cs, fuel-to-cladding mechanical interaction owing to the Cs-fuel-reactions by comparing the calculated results with post irradiation examinations.
Uwaba, Tomoyuki; Ohshima, Hiroyuki; Ito, Masahiro*
Nuclear Engineering and Design, 317, p.133 - 145, 2017/06
Times Cited Count：9 Percentile：65.76(Nuclear Science & Technology)The coupled numerical analysis of mechanical and thermal behaviors was performed for a wire-wrap fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal hydraulics analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that radial distribution of coolant temperatures in a subassembly tended to be flattened as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such temperature distribution was slightly analyzed as a result of the small bowing of the fuel pins due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal hydraulics was also investigated in this study.
Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko*; Imai, Yasutomo*; Ito, Masahiro*
AIP Conference Proceedings 1702, p.040011_1 - 040011_4, 2015/12
A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions including fuel deformation. This paper gives a summary of numerical methods of component programs of the system and their validation studies.
Uwaba, Tomoyuki; Mizuno, Tomoyasu; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*
Nuclear Engineering and Design, 280, p.27 - 36, 2014/12
Times Cited Count：11 Percentile：64.59(Nuclear Science & Technology)A deterministic computer code CEDAR has been developed to analyze irradiation behaviors of a mixed-oxide fuel pellet pin in a FBR. The FEM was incorporated into the mechanical calculation part of the code for properly analyzing stress-strain status in the fuel pellet and cladding, and mechanical interaction between the fuel pellet and cladding. The code features mechanistic analyses of irradiation behaviors of a fuel pin by integrating a lot of models to analyze major irradiation phenomena, thus expressing actual fuel pin irradiation behaviors. Analysis capabilities of the code were validated by calculations of fuel pellet temperatures, fractional fission gas releases of fuel pins and fuel pin cladding diametral strain profiles. The mechanisms of the fuel pin irradiation behaviors such as redistribution of Americium, PCMI and JOG formation were interpreted from the code analyses for the actual irradiation test fuel pins.
Uwaba, Tomoyuki; Ito, Masahiro*; Nemoto, Junichi*; Ichikawa, Shoichi; Katsuyama, Kozo
Journal of Nuclear Materials, 452(1-3), p.552 - 556, 2014/09
Times Cited Count：1 Percentile：8.88(Materials Science, Multidisciplinary)The BAMBOO code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle-duct interaction (BDI) condition. The pin diameters of were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype FBR and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT)images and local parameters of bundle deformation were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms.
Uwaba, Tomoyuki; Yano, Yasuhide; Ito, Masahiro*
Journal of Nuclear Materials, 421(1-3), p.132 - 139, 2012/02
Times Cited Count：7 Percentile：48.31(Materials Science, Multidisciplinary)Resistance spot welding of 11Cr-0.4Mo-2W,V,Nb ferritic/martensitic steel sheets with different thicknesses was examined to develop a manufacturing technology for a fast reactor fuel subassembly with an inner duct structure. In the spot welding, welding current, electrode force, welding time and holding time were varied as welding parameters to investigate the appropriate welding conditions. The formation of crack and void defects in the nugget could be suppressed by increasing the electrode force to 9.8 kN. It was also found that the electrode cap with a longer tip end length was effective for preventing weld defect formations. Strength of the spot welded joint was characterized from micro hardness and shear tension tests. In addition, the ductile-to-brittle transition temperature of the spot welded joint was measured by Charpy impact tests with specimens that had notches in the welded zone.
Uwaba, Tomoyuki; Ito, Masahiro*; Maeda, Koji
Journal of Nuclear Materials, 416(3), p.350 - 357, 2011/09
Times Cited Count：12 Percentile：66.82(Materials Science, Multidisciplinary)The C3M irradiation test, which was conducted in the experimental fast reactor in Joyo, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of 130 GWd/t safely. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. Contributions to the cladding irradiation creep strain were made by the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and 
Cs 
scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium urinate, which occurred near the top of the MOX fuel column, significantly increased the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.
Uwaba, Tomoyuki; Ito, Masahiro*; Mizuno, Tomoyasu; Katsuyama, Kozo; Makenas, B. J.*; Wootan, D. W.*; Carmack, J.*
Journal of Nuclear Materials, 412(3), p.294 - 300, 2011/05
Times Cited Count：11 Percentile：66.82(Materials Science, Multidisciplinary)The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence of about 39
10
n/m
as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.
Uwaba, Tomoyuki; Sogame, Motomu; Ito, Masahiro*; Mizuno, Tomoyasu; Donomae, Takako; Katsuyama, Kozo
Journal of Nuclear Science and Technology, 47(8), p.712 - 720, 2010/08
Times Cited Count：6 Percentile：40.28(Nuclear Science & Technology)In determining lifetime criteria of fast reactor fuel pins, creep damage due to fission gas pressure on mixed-oxide fuel pins with austenitic stainless steel cladding successfully irradiated to high burnups (120 GWd/t or higher pin averaged burnup) was evaluated. The degree of creep damage of these fuel pins was expressed as cumulative damage fractions (CDFs), defined so that cladding breaching occurs when the CDF exceeds 1.0. The obtained CDFs for typical high temperature fuel pins were on the order of 10
-10
at the end of irradiation, indicating that these fuel pins had large safety margins against breaching due to creep damage. In order to investigate the factors that govern the lifetime of fuel pins, pin diametral increase as well as CDF were predicted in cases of extended burnups from 120 GWd/t onward, and then were compared with tentatively determined limit values. The predicted pin diametral increase reached its limit value earlier than the CDF because of a significant increase in the cladding void swelling, suggesting that lifetimes of fuel pins with austenitic stainless steel cladding are practically governed by the diametral increase rather than by the creep damage.
Uwaba, Tomoyuki; Ito, Masahiro*; Mizuno, Tomoyasu
Journal of Nuclear Science and Technology, 45(11), p.1183 - 1192, 2008/11
Times Cited Count：13 Percentile：66.71(Nuclear Science & Technology)Uwaba, Tomoyuki; Ito, Masahiro*; Ukai, Shigeharu; Pelletier, M.*
Journal of Nuclear Science and Technology, 42(7), p.608 - 617, 2005/07
Times Cited Count：9 Percentile：53.26(Nuclear Science & Technology)The bundle-duct interaction analysis code "BAMBOO" incorporated models of the fuel pin self bowing induced by thermal and void swelling strain difference in the circumferential direction and pin array disarrangement called dispersion into the deformation analysis. The code was verified by the use of post irradiation examinations of two high burn-up subassemblies in Phenix reactor.
Ito, Masahiro*; Uwaba, Tomoyuki
JNC TN9400 2005-010, 72 Pages, 2005/04
JNC-TN9400-2005-010.pdf:3.54MB
JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code
Ito, Masahiro*; Imai, Yasutomo*; Uwaba, Tomoyuki; Ohshima, Hiroyuki
JNC TN9400 2004-003, 40 Pages, 2004/03
JNC-TN9400-2004-003.pdf:0.83MB
JNC has been developing a bundle deformation analysis code BAMBOO, a thermal hydraulics analysis code ASFRE and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupled analysis of deformation and thermal-hydraulics in a pin-bundle under a steady state condition just after startup for the purpose of the verification of the simulation system.
Uwaba, Tomoyuki; Ito, Masahiro*; Ohshima, Hiroyuki; Imai, Yasutomo*
JNC TN9400 2003-026, 132 Pages, 2003/04
JNC-TN9400-2003-026.pdf:6.16MB
A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBO0. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly.
Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa; Hashimoto, T.*; Koizumi, Koichi; *; *; *; Tada, Eisuke; Nakahira, Masataka; et al.
JAERI-Tech 95-019, 129 Pages, 1995/03
no abstracts in English
; Tone, Tatsuzo; ; ; ; ; ; ; ; ; et al.
JAERI-M 85-179, 454 Pages, 1986/01
no abstracts in English
