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Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Amaya, Masaki
JAEA-Data/Code 2013-014, 382 Pages, 2014/03
A light water reactor fuel analysis code FEMAXI-7 has been developed as the latest version for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. This report is the revised edition of the first one, JAEA-Data/Code 2010-035, which describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. The first edition was extended by orderly addition and disposition of explanations of models and organized as this revised edition after three years interval.
Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Nagase, Fumihisa
JAEA-Data/Code 2013-009, 306 Pages, 2013/10
A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7.
Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Nagase, Fumihisa
JAEA-Data/Code 2013-005, 382 Pages, 2013/07
A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions.
Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Nagase, Fumihisa
JAEA-Data/Code 2012-012, 374 Pages, 2012/07
A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which has been fully disclosed in the code model description published recently as JAEA-Data/Code 2010-035. The present manual, which is the counterpart of this description, gives detailed explanations of operation method of FEMAXI-7 code and its related codes, methods of input/output, methods of source code modification, features of subroutine modules, and internal variables in a specific manner in order to facilitate users to perform a fuel analysis with FEMAXI-7.
Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka
JAEA-Data/Code 2010-035, 361 Pages, 2011/03
A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior. This report describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions.
Teshigawara, Makoto; Harada, Masahide; Saito, Shigeru; Oikawa, Kenichi; Maekawa, Fujio; Futakawa, Masatoshi; Kikuchi, Kenji; Kato, Takashi; Ikeda, Yujiro; Naoe, Takashi*; et al.
Journal of Nuclear Materials, 356(1-3), p.300 - 307, 2006/09
Times Cited Count:9 Percentile:53.38(Materials Science, Multidisciplinary)We adopted silver-indium-cadmium (Ag-In-Cd) alloy as a material of decoupler for decoupled moderator in JSNS. However, from the heat removal and corrosion protection points of view, the Ag-In-Cd alloy is needed to clad between Al alloys (Al5083). We attempted to obtain good bonding conditions for between Al5083 and ternary Ag-In-Cd alloys by HIPing tests. The good HIP condition was found for small test piece (20mm). Though a hardened layer due to the formation of AlAg was found in the bonding layer, the rupture strength of the bonding layer was more than 20 MPa, which was the calculated design stress. Bonding tests of a large size piece (20020030 mm), which simulated the real scale, were also performed according to the results of small size tests. The result also gave good bonding and enough required-mechanical-strength, however the rupture strength of the large size test was smaller than that of small one.
Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 43(9), p.1097 - 1104, 2006/09
Times Cited Count:7 Percentile:46.04(Nuclear Science & Technology)The RANNS code analyzes behaviors of a single fuel rod in reactivity-initiated accident (RIA) conditions. The code has two types of mechanical model; one-dimensional deformation model for each axial segment length of rod, and newly-developed two-dimensional local deformation model for one pellet length. Analyses were performed on the RIA-simulated experiments in the Nuclear Safety Research Reactor (NSRR), OI-10, with high burnup PWR rods, and results of cladding deformation were compared between calculations by the two models and PIE data. The pre-accident, or End-of-Life conditions of the rods were predicted by the fuel performance code FEAMXI-6. In the calculations by the two-dimensional model of RANNS, the plastic strain increases at the cladding ridges during PCMI were compared with those in between the ridges and with the PIE data, and effect of stress variance induced by local non-uniformity of strain on the crack growth was discussed.
Suzuki, Motoe; Fuketa, Toyoshi; Saito, Hiroaki*
Nuclear Technology, 155(3), p.282 - 292, 2006/09
Times Cited Count:16 Percentile:72.85(Nuclear Science & Technology)Experimental analyses were performed for the RIA-simulated tests, OI-10 and OI-11 of high burnup PWR rods, in the NSRR by the RANNS code. The rod conditions were calculated by the fuel performance code FEMAXI-6 following the actual power history from the beginning to the end of irradiation in PWR and the results were given to the RANNS code as pre-test conditions. The RANNS analysis was conducted on the basis of such test conditions in the NSRR as the pre-test conditions, pulse power enthalpy and coolant temperature. The predicted quantities such as temperature of pellet stack and cladding, stress-strain distribution in cladding, and interactions among them during pulse irradiation were discussed in terms of PCMI process and compared with the experimental observations. In the OI-10 rod, calculated cladding permanent strain has a reasonable agreement with strain profile obtained in PIE, while locally enhanced strain of cladding was pointed out. In the OI-11 rod, the process from crack initiation to split failure was accounted for by the plastic strain occurrence in cladding.
Suzuki, Motoe; Saito, Hiroaki*
JAEA-Data/Code 2005-003, 415 Pages, 2006/02
A light water reactor fuel analysis code FEMAXI-6 is an advanced version which has been produced by integrating the former version FEMAXI-V with numerous functional improvements and extensions. In particular, the FEMAXI-6 code has attained a complete coupled solution of thermal analysis and mechanical analysis, enabling an accurate prediction of pellet-clad gap size and PCMI in high burnup fuel rods. Also, such new models have been implemented as pellet-clad bonding and fission gas bubble swelling, and linkage function with detailed burning analysis code has been enhanced. Furthermore, a number of new materials properties and parameters have been introduced. With these advancements, the FEMAXI-6 code has been upgraded to a versatile analytical tool for high burnup fuel behavior not only in the normal operation but also in anticipated transient conditions. This report describes in detail the design, basic theory and structure, models and numerical method, improvements and extensions, and method of model modification. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output of the code are also described, and a sample output in an actual form is included.
Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi
Nuclear Engineering and Design, 236(2), p.128 - 139, 2006/01
Times Cited Count:8 Percentile:49.58(Nuclear Science & Technology)A computer code RANNS was developed to analyze fuel rod behaviors in the RIA conditions. The code performs thermal and FEM-mechanical calculation for a single rod in axis-symmetric geometry to predict temperature profile, PCMI contact pressure, stress-strain distribution and their interactions. An experimental analysis by RANNS begins with pre-test conditions of irradiated rod which are given by FEMAXI-6. Analysis was performed on the simulated RIA experiments in NSRR, FK-10 and FK-12, of high burnup BWR rods in a cold start-up conditions, and PCMI process was discussed extensively. It was revealed that pellet thermal expansion dominates cladding deformation and subjects the cladding to bi-axial stress state, and thermal expansion in the cladding makes the stress in the inner region significantly lower than that in the outer region. Simulation calculations with wider pulses were carried out and the resulted cladding hoop stress was compared with the failure stress estimated in the NSRR experiments.
Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.579 - 601, 2005/10
The RANNS code analyzes behaviors of a single fuel rod in reactivity-initiated accident (RIA) conditions. The code has two types of mechanical model; one-dimensional deformation model for each axial segment length of rod, and newly-developed two-dimensional local deformation model for one pellet length. Analyses were performed on the two RIA-simulated experiments in the NSRR, OI-10 and OI-11 with high burnup PWR rods, and results of cladding deformation were compared between calculations by the two models and PIE data. RANNS calculated the deformation profiles of claddings during the power transient of the experiments on the basis of the pre-pulse conditions of rods predicted by FEMAXI-6 code. In the calculations by the two-dimensional model, the plastic strain increase at the cladding ridges was compared with those in between the ridges and with the PIE data, and effect of stress variance induced by local non-uniformity of strain on the crack growth was discussed.
Teshigawara, Makoto; Harada, Masahide; Saito, Shigeru; Kikuchi, Kenji; Kogawa, Hiroyuki; Ikeda, Yujiro; Kawai, Masayoshi*; Kurishita, Hiroaki*; Konashi, Kenji*
Journal of Nuclear Materials, 343(1-3), p.154 - 162, 2005/08
Times Cited Count:10 Percentile:56.74(Materials Science, Multidisciplinary)For decoupled and poisoned moderator, a thermal neutron absorber, i.e., decoupler, is located around the moderator to give neutron beam with a short decay time. A B4C decoupler is already utilized, however, it is difficult to use in a MW class source because of He void swelling and local heating by (n,a) reaction. Therefore, a Ag-In-Cd (AIC) alloy which gives energy-dependence of macroscopic neutron cross section like that of BC was chosen. However, from heat removal and corrosion protection points of view, AIC is needed to bond between an Al alloy (A6061-T6), which is the structural material of a moderator. An AIC plate is divided into a Ag-In (15wt%) and Ag-Cd (35wt%) plate to extend the life time, shorten by burn up of Cd. We performed bonding tests by HIP (Hot Isostatic Pressing). We found out that a better HIP condition was holding at 803 K, 100 MPa for 1 h for small test pieces (f20mm). Though a hardened layer is found in the bonding layer, the rupture strength of the bonding layer is more than 20 MPa, which is less than that of the design stress.
Suzuki, Motoe; Kusagaya, Kazuyuki*; Saito, Hiroaki*; Fuketa, Toyoshi
Journal of Nuclear Materials, 335(3), p.417 - 424, 2004/12
Times Cited Count:5 Percentile:35.25(Materials Science, Multidisciplinary)Experimental analysis was conducted on the Lift-Off experiment IFA-610.1 in Halden reactor by the FEMAXI-6 code using the detailed measured conditions of test-irradiation. Calculated fuel center temperatures on the two assumptions, i.e., (1) an enhanced thermal conductance across the pellet-clad bonding layer is maintained during the cladding creep-out by over-pressurization, and (2) the bonding layer is broken by the cladding creep-out, were compared with the measured data to analyze the effect of the creep-out by over-pressure inside the test pin. The measured center temperature rise was higher by a few tens of K than the prediction performed on the assumption (1), though this difference was much smaller than the predicted rise on the assumption (2). Therefore, it is appropriate to attribute the measured center temperature rise to the decrease of effective thermal conductance by irregular re-location of pellet fragments, etc. which was caused by cladding creep-out.
Suzuki, Motoe; Uetsuka, Hiroshi; Saito, Hiroaki*
Nuclear Engineering and Design, 229(1), p.1 - 14, 2004/04
Times Cited Count:19 Percentile:75.21(Nuclear Science & Technology)Mechanical load on cladding induced by fuel swelling in a high burn-up BWR type rod has been analyzed by a fuel performance code FEMAXI-6. The code has been developed for the analysis of LWR fuel rod behaviors in normal operation and transient conditions using FEM. During a power ramp for the high burn-up rod, instantaneous pellet swelling can significantly exceed the level that is predicted by a "steady-rate" swelling model, causing a large circumferential strain in cladding. This phenomenon has been simulated by a new swelling model to take into account the fission gas bubble growth, and as a result it has been found that the new model can give reasonable predictions on cladding diameter expansion in comparison with post-irradiation data. In addition, a pellet-clad bonding model which has been incorporated in the code to assume firm mechanical coupling between pellet outer surface and cladding inner surface has predicted the generation of bi-axial stress state in the cladding during ramp.
Suzuki, Motoe; Saito, Hiroaki*; Iwamura, Takamichi
Nuclear Engineering and Design, 227(1), p.19 - 27, 2004/01
Times Cited Count:7 Percentile:45.11(Nuclear Science & Technology)To assess the feasibility of the 31percentPu-MOX fuel rod design of reduced-moderation boiling water reactor in terms of thermal and mechanical behaviors, a single rod which is assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO fuel have been used complementally. The results are: FGR is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400K, while cladding diameter increase caused by pellet swelling is within 1percent strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling requires to be investigated in detail.
Suzuki, Motoe; Saito, Hiroaki*
JAERI-Data/Code 2003-019, 423 Pages, 2003/12
A light water reactor fuel analysis code FEMAXI-6 is an advanced version which has been produced by integrating the former version with a number of improvements. In particular, the FEMAXI-6 code has attained a complete coupled solution of thermal analysis and mechanical analysis, permitting an accurate prediction of pellet-clad gap size and PCMI in high burnup fuel rods. Also, such new models have been implemented as pellet-clad bonding and fission gas bubble swelling, and the coupling with burning analysis code has been enhanced. Furthermore, a number of new materials properties and parameters have been introduced. With these advancements, the FEMAXI-6 code is a versatile tool not only in the normal operation but also in transient conditions. This report describes the design, basic theory, models and numerical method, improvements, and model modification. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output, and a sample output in an actual form are included.
Saito, Hioraki*; Iriya, Yoshikazu*
JNC TJ8440 99-003, 156 Pages, 1999/03
no abstracts in English
Suzuki, Motoe; Saito, Hioraki*
HPR-349, 12 Pages, 1998/00
no abstracts in English