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Udagawa, Yutaka; Suzuki, Motoe; Amaya, Masaki
JAEA-Data/Code 2014-025, 27 Pages, 2015/02
JAEA-Data-Code-2014-025.pdf:2.53MB
A light water reactor fuel analysis code RANNS has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly reactivity-initiated accident (RIA) conditions. The recent model development for the RANNS code has been focused on improving predictability of stress, strain, and temperature inside a fuel rod during pellet cladding mechanical interaction (PCMI), which is one of the most important behaviors of high-burnup fuels under RIA conditions. This report provides descriptions of the models developed and/or validated recently via experimental analyses using the RANNS code on the RIA-simulating experiments conducted in the nuclear safety research reactor (NSRR): models for mechanical behaviors as relocation of fuel pellets, pellet yielding, pellet-cladding mechanical bonding, and PCMI failure limit of fuel cladding, and thermal behaviors as pellet-cladding gap conductance and heat transfer from fuel rod surface to coolant water.
Udagawa, Yutaka; Sugiyama, Tomoyuki; Suzuki, Motoe; Amaya, Masaki
IAEA-TECDOC-CD-1775; Proceedings of Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents (CD-ROM), p.200 - 219, 2015/00
Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Amaya, Masaki
JAEA-Data/Code 2013-014, 382 Pages, 2014/03
JAEA-Data-Code-2013-014.pdf:16.36MB
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.
Udagawa, Yutaka; Mihara, Takeshi; Sugiyama, Tomoyuki; Suzuki, Motoe; Amaya, Masaki
Journal of Nuclear Science and Technology, 51(2), p.208 - 219, 2014/02
Times Cited Count：9 Percentile：57.19(Nuclear Science & Technology)Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Nagase, Fumihisa
JAEA-Data/Code 2013-009, 306 Pages, 2013/10
JAEA-Data-Code-2013-009.pdf:5.73MB
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
JAEA-Data-Code-2013-005.pdf:6.4MB
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.
Udagawa, Yutaka; Sugiyama, Tomoyuki; Suzuki, Motoe; Nagase, Fumihisa
Journal of Nuclear Science and Technology, 50(6), p.645 - 653, 2013/06
Times Cited Count：8 Percentile：49.28(Nuclear Science & Technology)Akie, Hiroshi; Sato, Isamu; Suzuki, Motoe; Serizawa, Hiroyuki; Arai, Yasuo
Journal of Nuclear Science and Technology, 50(1), p.107 - 121, 2013/01
Times Cited Count：2 Percentile：18.63(Nuclear Science & Technology)A simple formula is developed for the evaluation of the helium production amount in the fast reactor fuel. For the subroutine use in the existing fuel behavior analysis code, the formula is designed putting emphasis on simplicity and quickness rather than accuracy. The accuracy of the formula is confirmed by comparing with the detailed calculation with SWAT code, and also with the post irradiation examination (PIE) results of the fuel pin irradiated at the experimental fast reactor JOYO. As a result, the formula is found to evaluate the helium amount with the difference of less than about 10% from the detailed calculation and from the PIE results. Based on these results, the formula is installed in the fuel behavior analysis code for the simulation of helium behavior in fast reactor fuels.
Ogiyanagi, Jin; Hanawa, Satoshi; Suzuki, Motoe; Nagase, Fumihisa
Nuclear Engineering and Design, 253, p.77 - 85, 2012/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Irradiation behavior of medium and high burnup BWR fuels during base-irradiation and power ramp test is analyzed by a fuel performance code FEMAXI-7. The calculated values such as fission gas release after the base-irradiation and a cladding diameter profile before and after the ramp test show a reasonable agreement with measured PIE data. It was also found that the code can reasonably predict the FGR at high temperature condition up to 1800
C of pellet center temperature by using the FGR model indigenous to the code and the enhanced value of the original Turnbull model of fission gas atoms diffusion constant. For the ridging deformation of the cladding before and after the ramp test, the local PCMI analysis with 2-D geometry in FEMAXI-7 gave a reasonable agreement with the PIE data. Thus, it is demonstrated that the FEMAXI-7 code can give an appropriate insight into the complicated thermal and mechanical interactions in medium and high burnup BWR fuel rods.
Suzuki, Motoe; Udagawa, Yutaka; Sugiyama, Tomoyuki; Nagase, Fumihisa
Proceedings of Annual Topical Meeting on Water Reactor Fuel Performance (TopFuel 2012) (USB Flash Drive), 6 Pages, 2012/09
Behavior of fission gas release (FGR) analysis is performed for the high burnup PWR fuels which are pulse-irradiated in the simulated Reactivity-Initiated Accident (RIA) experiment conducted at NSRR (Nuclear Safety Research Reactor) in Japan Atomic Energy Agency. The FGR model consists of two main parts: FEMAXI-7 calculates fission gas bubble growth in grain boundaries during base-irradiation, while RANNS performed the grain separation and burst release of gas from grain boundary inventory in the rapidly heated pellet in the RIA experiment. The calculated results are compared with the measured data, which resulted in a rough agreement for the amount of fission gas in pellet and burst FGR.
Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Nagase, Fumihisa
JAEA-Data/Code 2012-012, 374 Pages, 2012/07
JAEA-Data-Code-2012-012.pdf:5.39MB
JAEA-Data-Code-2012-012-appendix(CD-ROM).zip:0.08MB
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.
Hanawa, Satoshi; Ogiyanagi, Jin; Suzuki, Motoe
Journal of Nuclear Science and Technology, 49(5), p.516 - 525, 2012/05
Times Cited Count：2 Percentile：17.8(Nuclear Science & Technology)He pressurization effect on fission gas release (FGR) of BWR fuel rods under power transient conditions was analyzed by the fuel performance code FEMAXI-7. The experimental data provided to this study was obtained in the Halden reactor. Two rods were irradiated in the Halden reactor for 12 years in the IFA-409 as base-irradiation, then provided to the IFA-535 for power ramp tests to understand He-pressurization effect on fission gas release under power transient conditions, by adjusting internal pressure of the rods before power ramp test. FEMAXI-7 reasonably reproduced the experimental data of cladding elongation change and FGR behavior during the power ramp test. Based on the calculation results, the cause that apparent He-pressurization effect was not observed in the experiment was considered to be caused by insufficient gas communication during strong PCMI and gap thermal conductance by the solid-solid contact due to gap closure.
Arai, Yasuo; Serizawa, Hiroyuki; Nakajima, Kunihisa; Takano, Masahide; Sato, Isamu; Katsuyama, Kozo; Akie, Hiroshi; Suzuki, Motoe; Shirasu, Noriko; Haga, Yoshinori; et al.
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
High amount of He is generated in MA-bearing fuel during irradiation and storage periods compared with that in U or U-Pu fuel. Laboratory scale experiments, post irradiation examinations and modeling study were carried out in order to understand the He behavior in MA-bearing oxide fuel. Diffusion characteristics of He in single-crystal UO
were investigated by the Knudsen effusion mass spectrometry. Effects of the He accumulation on lattice and bulk expansion of oxide pellets were examined by use of alpha-decay of 
Cm. Post irradiation examinations of 0.5%Am-MOX fuel irradiated at a fast test reactor JOYO were carried out, concentrating on the He behavior in the fuel pellets. A model describing the He behavior in MA-MOX fuel was constructed based on the principle processes, such as generation, diffusion, equilibrium and release to outer gaseous phase. By use of the model as a subroutine of a conventional fuel behavior analysis code, the He behavior in MA-MOX fuel for fast reactors was simulated.
Suzuki, Motoe; Nagase, Fumihisa
Proceedings of 2011 Water Reactor Fuel Performance Meeting (WRFPM 2011) (CD-ROM), 7 Pages, 2011/09
The behavior of MOX fuels which were base-irradiated in PWR and test-irradiated in the Halden reactor was analyzed by the latest version of fuel performance code FEMAXI-7. For the calculation conditions, linear heat rate history, power density profile, and coolant condition etc. were given consistently from the base- to test-irradiation to predict the fuel temperature, fission gas release rate, and cladding deformation, etc. Comparison of calculated values with the measured data during the test-irradiation shows a reasonable agreement in thermal analysis results such as fuel temperatures and fission gas release rates, while the cladding deformation, which is involved with various interactions, suggests that it is still necessary to analyze and consider an optimum combination of models and their parameters to obtain a satisfactory prediction.
Sugiyama, Tomoyuki; Udagawa, Yutaka; Suzuki, Motoe; Nagase, Fumihisa
Proceedings of 2011 Water Reactor Fuel Performance Meeting (WRFPM 2011) (CD-ROM), 6 Pages, 2011/09
The Japan Atomic Energy Agency has performed pulse irradiation tests using the NSRR to investigate fuel behavior under Reactivity-Initiated Accident (RIA) conditions. The NSRR tests have provided data of the pellet-cladding mechanical interaction (PCMI) failure of high burnup fuels up to 77 GWd/t. In particular, the PCMI failure limit is the important information which is needed in the reactor safety review. However, there are some differences between the NSRR tests and RIAs supposed in power reactors, such as the coolant temperature and the width of power pulse. Influence of these differences should be quantitatively evaluated in order to estimate the PCMI failure limit anticipated under the power reactor conditions from the NSRR data. This paper presents experimental results from a set of room and high temperature RIA tests, and discusses the evaluation procedure of the influence of coolant temperature and power pulse width on the failure limit on the basis of the experimental data.
Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka
JAEA-Data/Code 2010-035, 361 Pages, 2011/03
JAEA-Data-Code-2010-035.pdf:4.25MB
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.
Yamaji, Akifumi; Suzuki, Motoe; Okubo, Tsutomu
JAEA-Research 2010-029, 54 Pages, 2010/09
JAEA-Research-2010-029.pdf:3.07MB
This study has evaluated uncertainties in FEMAXI-6 calculations and clarified key models and parameters for predicting LWR MOX fuel rod behavior. Irradiation data obtained from the Halden reactor experiments (IFA-597.4 rod-10, rod-11, and IFA-514 rod-1) were used. The maximum discharge burnup was about 40 GWd/tMOX (IFA-514 rod-1). The results showed that uncertainties in fission gas release calculations were particularly high, and contributions of pellet relocation, densification and swelling models on pellet temperature were also evaluated. The basic fission gas release mechanism of MOX fuels should be the same as that of UO fuels, but the parameters in the model need to be revised for MOX fuels. More experimental data are needed. However, frequent reactor shutdowns and restarts may cause pellet relocation changes which need to be considered in the evaluations.
Yamaji, Akifumi; Suzuki, Motoe; Okubo, Tsutomu
Journal of Nuclear Science and Technology, 46(12), p.1152 - 1161, 2009/12
Times Cited Count：5 Percentile：35.97(Nuclear Science & Technology)The advanced reactor concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been proposed and being studied to achieve effective and flexible utilization of the uranium and plutonium resources based on the well-developed light water reactor (LWR) technology. In order to design and evaluate the FLWR fuel rod behavior, the uncertainties in the FEMAXI-6 calculations and the key models and parameters for predicting LWR MOX fuel rod behavior need to be evaluated. In this study, the Test-Fuel-Data-Base (TFDB) obtained from the Halden reactor experiments (IFA-597.4 rod-10, rod-11, and IFA-514 rod-1) were used for the evaluations. The maximum discharge burnup was about 40 GWd/tMOX.
Udagawa, Yutaka; Suzuki, Motoe; Sugiyama, Tomoyuki; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 46(10), p.1012 - 1021, 2009/10
Times Cited Count：7 Percentile：45.28(Nuclear Science & Technology)Suzuki, Motoe; Sugiyama, Tomoyuki; Udagawa, Yutaka; Nagase, Fumihisa; Fuketa, Toyoshi
Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 11 Pages, 2009/09
The fuel behavior during the fast transients in the two cases of NSRR experiments using high burnup PWR fuel rods are analyzed by using the RANNS code. In one case, cladding failure occurred whereas in the other case the rod survived but gave rise to departure from nucleate boiling. The analytical results are compared with the metallographic observations of failed part of the cladding to discuss the failure-determining condition in terms of incipient crack depth, temperature and stress at the depth. Based on these evaluations, sensitivity study with respect to the effect of initial temperature on the stress/strain of cladding is conducted. In addition, simulations are performed in commercial reactor conditions. The results were compared with each other and failure capability of cladding is discussed.
