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Udagawa, Yutaka; Tasaki, Yudai
JAEA-Data/Code 2021-007, 56 Pages, 2021/07
JAEA-Data-Code-2021-007.pdf:5.05MB
Japan Atomic Energy Agency (JAEA) has released FEMAXI-8 in 2019 as the latest version of the fuel performance code FEMAXI, which has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly normal operation conditions and anticipated transient conditions. This report summarizes a newly developed model to analyze intragranular fission gas behaviors considering size distribution of gas bubbles and their dynamics. While the intragranular fission gas behavior models implemented in the previous FEMAXI versions have supported only treating single bubble size for a given fuel element, the new model supports multiple gas groups according to their size and treats their dynamic behaviors, making the code more versatile. The model was first implemented as a general module that takes arbitrary number of bubble groups and spatial mesh division for a given fuel grain system. An interface module to connect the model to FEMAXI-8 was then developed so that it works as a 2 bubble groups model, which is the minimum configuration of the multi-grouped model to be operated with FEMAXI-8 at the minimum calculation cost. FEMAXI-8 with the new intragranular model was subjected to a systematic validation calculation against 144 irradiation test cases and recommended values for model parameters were determined so that the code makes reasonable predictions in terms of fuel center temperature, fission gas release, etc. under steady-state and power ramp conditions.
Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12
Times Cited Count：5 Percentile：58.6(Nuclear Science & Technology)no abstracts in English
Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06
Times Cited Count：9 Percentile：73.24(Nuclear Science & Technology)no abstracts in English
Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki
JAEA-Data/Code 2018-016, 79 Pages, 2019/01
JAEA-Data-Code-2018-016.pdf:2.75MB
FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.
Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1183 - 1189, 2016/04
Suzuki, Motoe; Fuketa, Toyoshi; Saito, Hiroaki*
Nuclear Technology, 155(3), p.282 - 292, 2006/09
Times Cited Count：16 Percentile：73.32(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*; Fuketa, Toyoshi
Nuclear Engineering and Design, 236(2), p.128 - 139, 2006/01
Times Cited Count：8 Percentile：49.94(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.
Sugiyama, Tomoyuki; Nagase, Fumihisa; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.912 - 932, 2005/10
High burnup fuel cladding can fail due to mechanical interaction with expanding fuel pellet under reactivity initiated accident (RIA) conditions. In order to evaluate the cladding failure limit, investigations to modify ring tensile test have been performed to measure mechanical properties of Zircaloy cladding properly. JAERI developed the test method and geometry minimizing undesirable effects of friction and bending moment in the specimen. Using the modified test method, mechanical properties of unirradiated Zircaloy-4 cladding were evaluated as functions of hydrogen concentration and temperature. For hydrogen concentrations above 700 ppm, obvious increase of ductility is observed with the temperature increase from 300 to 473 K. For hydrogen concentrations below 500 ppm, on the other hand, temperature dependence of ductility is relatively small in the present temperature range from 300 to 573 K.
Fuketa, Toyoshi; Sugiyama, Tomoyuki; Sasajima, Hideo; Nagase, Fumihisa
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.633 - 645, 2005/10
LWR fuel behaviors during a reactivity initiated accident (RIA) are being studied in the NSRR program. Results from recent NSRR experiments, no failures in Tests OI-10 and -12 and the higher failure enthalpy in Test OI-11, reflect the better performance of the new cladding materials in terms of corrosion during PWR operations. Accordingly, these rods with improved corrosion resistance have larger safety margin than conventional Zircaloy-4 rods. In addition, the smaller inventory of inter-granular gas in the large grain pellet could reduce the fission gas release in RIA as observed in the OI-10. Test VA-1 was conducted with an MDA sheathed 78 MWd/kgU PWR fuel rod. Despite of the higher burnup and thicker oxide layer of 
81
m, the enthalpy at failure remained in a same level as those for rods with of 40m-oxide at 50 - 60 MWd/kgU. This result suggests high burnup structure (rim structure) in pellet periphery does not have strong effect on the failure enthalpy reduction because the PCMI load is produced primarily by solid thermal expansion of the pellet.
Tomiyasu, Kunihiko; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi
JAERI-Research 2005-022, 128 Pages, 2005/09
JAERI-Research-2005-022.pdf:16.11MB
In order to clarify the driving force of PCMI failure on high burnup fuels and the influence of hydrogen embrittlement on failure limit under RIA conditions, simulated-RIA experiments were performed on fresh fuel rods in the NSRR. The driving force was restricted only to thermal expansion of pellet by using fresh pellets, and fresh claddings were pre-hydrided to simulate hydrogen absorption of high burnup fuels. In seven experiments, test rods resulted in PCMI failure, which was observed on high burnup fuels, in terms of transient behavior and fracture configuration. It indicates that the driving force is sufficiently explained with thermal expansion of pellet and a contribution of fission gas is small. Many incipient cracks were generated in the outer surface of the cladding, and they stopped at the boundary between hydride rim and metallic layer. It suggests that a toughness of metallic region except hydride rim has particular importantance for failure limit. Fuel enthalpy at failure correlates with the thickness of hydride rim, and tends to decrease with thicker hydride rim.
Sugiyama, Tomoyuki; Fuketa, Toyoshi; Ozawa, Masaaki*; Nagase, Fumihisa
Proceedings of 2004 International Meeting on LWR Fuel Performance, p.544 - 550, 2004/09
Two pulse irradiation experiments simulating reactivity initiated accidents were performed on high burnup (60 GWd/t) PWR UO
rods with advanced cladding alloys. Test OI-10 was performed on an MDA cladded rod with large-grain (25 m) fuel pellets with a peak fuel enthalpy condition of 435 J/g, and resulted in a peak residual hoop strain of 0.7%. On the other hand, Test OI-11 on a ZIRLO cladded rod with conventional pellets resulted in a fuel failure at a fuel enthalpy of 500 J/g due to the pellet-cladding mechanical interaction (PCMI). A long axial split was generated on the cladding over the active length. The fuel pellets were fragmented and dispersed into the coolant water. The fuel enthalpy at failure is higher than the PCMI failure criterion of 209 J/g at the corresponding burnup. The experimental results suggest that the rods with improved corrosion resistance have much safety margin against the PCMI failure compared to the conventional Zircaloy-4 rod.
Fuketa, Toyoshi; Sugiyama, Tomoyuki; Nakamura, Takehiko; Sasajima, Hideo; Nagase, Fumihisa
NUREG/CP-0185, p.161 - 172, 2004/00
To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) is being studied in the Nuclear Safety Research Reactor (NSRR) program of the Japan Atomic Energy Research Institute (JAERI). A series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the NSRR power burst experiments with irradiated PWR fuels with ZIRLO and MDA claddings, and discusses effects of pellet expansion as PCMI (Pellet-Cladding Mechanical Interaction) loading and cladding embrittlement primarily due to hydrogen absorption. Separate-effect studies including tube-burst and ring-tensile tests on Zircaloy cladding also described.
Suzuki, Motoe; Saito, Hiroaki*
JAERI-Data/Code 2003-019, 423 Pages, 2003/12
JAERI-Data-Code-2003-019.pdf:17.7MB
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.
Suzuki, Motoe; Uetsuka, Hiroshi
IAEA-TECDOC-CD-1345 (CD-ROM), p.217 - 238, 2003/03
A fuel performance code FEMAXI-6 has been developed for the analysis of LWR fuel rod behaviors. The code uses FEM analysis, and has incorporated thermal and mechanical models of phenomena anticipated in high burn-up fuel rods. In the present study, PCMI induced by swelling in a high burn-up BWR type fuel rod has been analyzed. During a power ramp for the high burn-up fuel, instantaneous pellet swelling been simulated by a new swelling model which has been installed in the code to take into account the FP gas bubble growth, and the new model can give satisfactory predictions on cladding diametral expansion. In addition, a pellet-clad bonding model in the code, which assumes firm mechanical coupling between pellet outer surface and cladding inner surface, predicts an increased tensile stress in the axial direction of cladding during the power ramp, indicating the generation of bi-axial stress state in the cladding.
Nakamura, Jinichi; Nakamura, Takehiko; Sasajima, Hideo; Suzuki, Motoe; Uetsuka, Hiroshi
HPR-359, Vol.2, p.34_1 - 34_16, 2002/09
In BWR, power oscillations can occur due to the void fraction fluctuation. To investigate the fuel behavior during power oscillation of BWRs, two types of irradiated fuel rods were tested under simulated power oscillation conditions in the Nuclear Safety Research Reactor(NSRR). One is high burnup BWR fuel (56GWd/t) test, with 4 power oscillation cycles, to clarify the behavior of high burnup fuel. The second one is high enriched fuel(20%,25GWd/t) test, with 7 power cycles, to perform the test under high power conditions. The fuel behavior data, such as cladding elongation, fuel stack elongation, cladding temperature, etc. were obtained in these tests. The DNB did not occur in these tests. The PCI was observed through cladding elongation and fuel stack elongation during the power oscillations, but the residual strain of cladding was very small. Fuel behavior under simulated power oscillations is discussed based on in-pile data and PIE data and is compared with FEMAXI-6 and FRAP-T6 calculation.
61GWd/tU) BWR fuel behavior under reactivity initiated accident conditionsNakamura, Takehiko; Kusagaya, Kazuyuki*; Yoshinaga, Makio; Uetsuka, Hiroshi
JAERI-Research 2001-054, 49 Pages, 2001/12
JAERI-Research-2001-054.pdf:6.7MB
no abstracts in English
Boyack, B. E.*; Motta, A. T.*; Peddicord, K. L.*; Alexander, C. A.*; Deveney, R. C.*; Dunn, B. M.*; Fuketa, Toyoshi; Higar, K. E.*; Hochreiter, L. E.*; Langenbuch, S.*; et al.
NUREG/CR-6742, 263 Pages, 2001/09
no abstracts in English
Suzuki, Motoe
Saishin Kaku Nenryo Kogaku; Kodoka No Genjo To Tembo, p.131 - 140, 2001/06
no abstracts in English
Kitano, Koji*; Fuketa, Toyoshi; Sasajima, Hideo; Uetsuka, Hiroshi
JAERI-Research 2001-011, 34 Pages, 2001/03
JAERI-Research-2001-011.pdf:2.48MB
no abstracts in English
